NUCLEAR OPTION IN COUNTRIES WITH SMALL AND MEDIUM ...

Croatian Nuclear Society

in cooperation with

10th International Conference on

N U C L E A R O P T I O N IN COUNTRIES WITH SMALL AND MEDIUM ELECTR IC ITY G R I DSJune 1-4, 2014, Zadar, Croatia

B o o k o f A b s t r a c t s

IAEA

Editors

Nikola Čavlina, Davor Grgić, Dubravko Pevec

Edited and printed by

Croatian Nuclear SocietyUnska 3, 10000 Zagreb, Croatiatel: +385 1 612 9627, fax: +385 1 612 9605e-mail: [email protected]: www.nuclear-option.org, www.nuklearno-drustvo.hr

Editorial note

The views expressed in the papers, the statements made and the general styles adopted are the responsibility of the named authors. Authors are themselves responsible for obtaining the necessary permission to reproduce copyright material from other sources.

ISBN 978-953-55224-7-8

A CIP catalogue record for this book is available from the National and University Library in Zagreb under 878833.

10th International Conference on

N U C L E A R O P T I O N IN COUNTRIES WITH SMALL AND MEDIUM ELECTR IC ITY G R I DSJune 1-4, 2014, Zadar, Croatia

B o o k o f A b s t r a c t s

Conference Organized by

Croatian Nuclear Society in cooperation with IAEA,

Croatian State Office for Radiological and Nuclear Safety and University of Zagreb, Faculty of Electrical Engineering and Computing

Under the Auspices of

Ministry of Economy, Labour and Entrepreneurshup

Croatian Chamber ofEconomy

IRB

EUROPEAN NUCLEAR SOCIETY

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Programme CommitteeN. Čavlina, Chairman, Croatia

Josip Lebegner, HEP, CroatiaVladimir Lokner, APO, CroatiaNiko Malbaša, EKONERG, CroatiaGordana Marović, IMI, CroatiaBorut Mavko, IJS, SloveniaSaša Medaković, DZRNS, CroatiaJozef Mišák, REZ, Czech RepublicMichael Modro, IAEA, USABojan Niceno, PSI, SwitzerlandHisashi Ninokata, POLIMI, ItalyHrvoje Perharić, NEK, SlovenijaBojan Petrović, GATECH, USAAlessandro Petruzzi, UNIPI, ItalyDubravko Pevec, FER, CroatiaIvan Poljičanin, FOND – NEK,

CroatiaNik Popov, FYR MacedoniaAndrej Prošek, IJS, SloveniaFrederik Reitsma, IAEA, Austria S. M. S. Reitsma, SWISS RE,

Switzerland

Chris Allison, ISS, USATomislav Bajs, ENCONET, CroatiaIvica Bašić, APOS, CroatiaMikhail Bykov, GIDROPRESS, RussiaGerard Cognet, CEA, FranceDuško Čorak, INETEC, CroatiaFrancesco D’Auria, UNIPI, ItalyVesna Dimitrijević, AREVA, USADanilo Feretić, FER, CroatiaAndreas Fristedt-Ablad,

WESTINGHOUSE, SwedenMichel Giot, UNI-LOUVAIN,

BelgiumDavor Grgić, FER, CroatiaZoran Heruc, NEK, SloveniaTomasz Jackowski, UCBJ, PolandIgor Jenčič, IJS, SloveniaDarko Kavšek, NEK, SloveniaVladimir Knapp, FER, CroatiaMarjan Kromar, IJS, SloveniaVladimir Kuznetsov, IAEA, Austria

Nikola Rendić-Miočević, HR NUKPOOL, Croatia

Francesc Reventos, UPC, SpainMarco Ricotti, POLIMI, ItalyStane Rožman, NEK, SloveniaRainer Salomaa, UNI-AALTO, FinlandVladimir Slugen, STUBA, SlovakiaZoran Stošić, AREVA, GermanyZdenko Šimić, FER, CroatiaPredrag Širola, NEK, SloveniaTonči Tadić, IRB, CroatiaŽeljko Tomšić, FER, CroatiaCsilla Toth, PAKS2, HungaryAndrej Trkov, IJS, SloveniaEugenijus Uspuras, LEI, LithuaniaJean-Pierre Van-Dorsselaere, IRSN,

FranceKiril Velkov, GRS, GermanyIvan Vrbanić, APOS, CroatiaMladen Zeljko, EIHP, CroatiaTomaž Žagar, ARAO, Slovenia

Organising CommitteeD. Pevec, Chairman, Croatia

Luka Alujević, HEP, CroatiaAna Holjak, FER, Croatia

Irena Jakić, HEP, CroatiaSiniša Šadek, FER, Croatia

Nikola Škorić, NEK, SloveniaKrešimir Trontl, FER, Croatia

10TH INTERNATIONAL CONFERENCE: NUCLEAR OPTION IN COUNTRIES WITH SMALL AND MEDIUM ELECTRICITY GRIDS

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1 – 4 June 2014, Zadar, Croatia

Foreword

The International Conference ”Nuclear Option in Countries with Small and Medium Electricity Grids” is the tenth in a series of meetings on the same topics organized biennially by the Croatian Nuclear Society. This topical conference was initiated in 1996 and the first conference took place in Opatija, the following seven in Dubrovnik and the last one in Zadar. This year, it again takes place in Zadar. The conference is organized with intention to focus on specific aspects of usage of nuclear energy for electricity production in small and medium sized countries.Importance of international cooperation for the assessment of the nuclear option has been recognised by the International Atomic Energy Agency (IAEA). As a result of this recognition, the Conference is organized in co-operation with IAEA. Croatian State Office for Radiological and Nuclear Safety and Uni-versity of Zagreb, Faculty of Electrical Engineering and Computing have also participated in Conference organization.Session topics reflect some current emphasis, such as country energy needs, new reactor technologies, operation and safety of the operating nuclear power plants. The conference also focuses on the exchange of experience and co-operation in the fields of fuel cycle, radioactive waste management, regulatory prac-tice and liability.Authors’ contributions are presented in 12 invited and 61 contributed papers. All the contributed papers are grouped in 8 sessions:

1. Energy Planning and Nuclear Option2. Power Reactors and Technologies3. Nuclear Safety4. Radioactive Waste Management and Decommissioning5. Operation and Maintenance Experience6. Environment, Public Relations and Safety Culture7. Regulatory Practice and Emergency Preparedness8. Reactor Physics and Nuclear Fuel Cycle

All contributed papers are published. The abstracts are printed in the Book of abstracts, and Proceedings with full papers are given on the USB memory stick.We would like to express our gratitude to over 100 authors and co-authors that put an extra effort into completing their full camera-ready papers. We would also like to thank the sessions’ coordinators, re-viewers and all those who have helped us in organizing this Conference.

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Acknowledgements are given to the International Atomic Agency and the European Nuclear Society for their support and encouragement.Also, we are particularly grateful to all the sponsors whose contribution has been essential for the suc-cess of this International Conference. We express our thanks to all those, who through their efforts and participation, have contributed to the Conference success.

Zagreb, May 2014Editors


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Conference Sponsored by

CROATIANNUCLEAR

POOLA SPO

TECHNA

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Invited lectures

IL-01 S. M. ModroThe IAEA Safety Standards and SMR design features 19

IL-02 F. RomanelliThe European Roadmap to Fusion Electricity 20

IL-03 N. ČavlinaAssessment of Energy Options for CO2 Emission Reduction 21

IL-04 N. ShulyakWestinghouse Small Modular Reactor (SMR) Program 22

IL-05 N. ShulyakWestinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs 24

IL-06 Z. V. StošićReactors Project Delivery: The Value of Experience 25

IL-07 B. PetrovićProgress in Development of the I2S-LWR Concept 26

IL-08 I. SandaSafety Assessment Challenges of the New Generation Reactors – Experience from the IAEA Generic Reactor Safety Reviews 27

IL-09 J. ŠpilerNuclear Power Plant Krško 2 Project Planning and Preparation Phase 28

IL-10 C. AllisonUpdate on the SDTP sponsored Fukushima Daiichi Related Assessment Activities 29

Table of Contents


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IL-11 T. ŽagarRadwaste Management in Small Nuclear Country – National Policy and Strategy 30

IL-12 H. ZaccaiAREVA “Sustainable Cycle Solutions” 31

Session 1 Energy Planning and Nuclear Option

S1-148 V. Knapp, D. PevecA View on the Future of Nuclear Fission Energy 34

S1-149 V. Knapp, D. Pevec, M. Matijević, D. LaleMolten Salt Thorium Reactor – A Promising Nuclear Technology to Stop Global Warming 35

S1-162 Ž. TomšićFrom Market Uncertainty to Policy Uncertainty for Investment in Power Generation: Real Options for NPP on Electricity Market 36

Session 2 Power Reactors and Technologies

S2-106 M. Kim, S. Hong, Y. Kim, D. Shin, J. Lee, G. ParkStatus of Thermo-Fluid Experimental Research on VHTR in Korea 38

S2-116 H. Grganić, B. GlaserComparison of AP1000 and Generation II PWR with an Accent on Safety Systems and Coping with Station Blackout 39

S2-141 B. Niceno, M. PouchonThermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides 40

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S2-151 D. Grgić, T., Fancev, V. Benčik, S. ŠadekHot Leg Streaming Calculation for Two-Loop PWR Plant 41

S2-216 A. Manera, M. J. MemmottDesign and Trade-off Studies of the Passive Decay Heat Removal System (DHRS) of the Integral Inherently Safe LWR (I2S-LWR) 42

Session 3 Nuclear Safety

S3- 101 A. ProšekSensitivity analysis of LOFT L2-5 test calculations 45

S3- 117 G. Vayssier, I. BašićSystematic Review of Accident Management Programs – Principles, Experiences 46

S3-121 B. MarinovaSeismic margin assessment for nuclear facilities of Kozloduy NPP 47

S3-125 I. Horvatovic, C. Batra, M. Cherubini, A. Petruzzi, F. D’auria, T. Bajs Benchmark analysis of EBR–II shutdown heat removal test SHRT–17 48

S3- 135 M. Uršič, M. Leskovar, R. MeignenAnalysis of Effect of Sodium Thermo-Dynamical Properties in Fuel-Coolant Interaction 49

S3- 137 M. Mihalina, S. Špalj, B. Glaser, R. Jalovec, G. JankovićNPP Krško SAMG Upgrade 50

S3- 139 S. Šadek, D. GrgićContainment Modelling with the ASTEC Code 51

S3- 146 K. ManchevaMain Aspects and Results of Level 2 PSA for KNPP WWER-1000/B320 52


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S3- 150 D. Grgić, T. Fancev, S. ŠadekCalculation of hydrogen concentration in containment during LOCA accident 53

S3- 153 V. Benčik, D. Grgić, N. Čavlina, S. ŠadekOptimization of OPDT Protection for Overcooling Accidents 54

S3- 157 R. Banov, Z. Šimić, D. ŠtercShort overview of PSA quantification methods, pitfalls on the road from approximate to exact results 55

S3- 163 D. Slovenc, I. Bašić, L. ŠtrubeljSimulation and validation of a full and mini flow surveillance test in NEK with best estimate HPIS model in APROS 56

Session 4 Radioactive Waste Management and Decommissioning

S4-112 O. Purtov, A. Masko, V. VasilchenkoMain Results of Updated Decommission Conception of NPPs Operating in Ukraine 58

S4-132 A. Knapp, I. Levanat, D. Šaponja-MilutinovićIs Croatia Going to Build a Radioactive Waste Repository? 59

S4-133 I. Levanat, V. LoknerCroatian Capacity for Management of the NPP Krsko Radioactive Waste 60

S4-134 L. AlujevićThe Safety Assessment Framework Tool (SAFRAN) – Description, Overview and Applicability 61

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Session 5 Operation and Maintenance Experience

S5-102 T. Tomašić, I. Vuković, A. BakićARCHER – Advanced System for RPVH Inspection and Repair 64

S5-103 M. Vavrous, K. Markulin, M. BrekaloIn-service inspection of primary circuit components: Qualification of non-destructive testing applications 65

S5-104 M. Kekelj, M. Budimir, N. Pavlović, R. GracinMultielement Ultrasound and Eddy Current Integrated Probe for Non-destructive Evaluation of Nuclear Reactor Pressure Vessel Head Penetrations 66

S5-105 P. MateljakCASTOR – Advanced System for VVER Steam Generator Inspection 67

S5-107 D. Borović, I. VukovićGIMIS – Integral Solution for the In-Service Inspection Management of Components in NPPs 68

S5-118 A. Bakić, M. Pajnić, T. TomašićNew Solution For Ultrasonic Pipe Inspection System 69

S5-124 Z. Šimić, R. BanovDevelopment of the Operational Events Groups Ranking Tool 70

S5-129 S. ReškovićAutomatic Analysis for VVER and PWR 71

S5-136 B. Bach, R. Čižmek, B. BožinDeveloping Effective Corrective Action Plan in Krško NPP 72

S5-138 M. Chambers, B. Kurinčić16x16 Vantage+ Fuel Assembly Flow Vibrational Testing 73

S5-140 D. Djaković


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Control of Special Processes at Krško NPP 74

S5-142 I. Vrbanić, I. BašićOn Use of PSA for Characterization of Risk Significance of Operational Events and Issues in NPPs 75

S5-161 M. Gluhak, M. SenegovićShutdown Safety in NEK 76

S5-206 M. DudašOperator Fundamentals 77

Session 6 Environment, Public Relations and Safety Culture

S6-113 B. Manchev, V. Yordanova, B. NenkovaConfiguration Management Program – a part of Integrated Management System 79

S6-115 K. Trontl, D. Pevec, M. Matijević, R. Ječmenica, J. LebegnerPublic Opinion Survey – Energy – The Present and the Future – 2012/2013 81

S6-123 R. Bišćan, I. FifnjaKrško NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System) 82

S6-131 I. Jakić, R. FilipinAnalysis of Public Opinion Survey “Nuclear Energy – the Present and the Future” (2000–2012) 84

S6-147 S. Pleslić, G. Jimenez VarasKnowledge Loss Risk Assessment in Education and Industry 85

S6-158 I. Prlić, M. Surić Mihić, P. Shaw, M. Hajdinjak, Ž. Božina, D. Kosmina, Z. Cerovac

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EU Outer Borders and Radiation – An urgent need for standardization, new detector technologies and education harmonization 86

S6-159 I. Prlić, M. HajdinjakNORM – Radionuclide transfer studies- A modern approach according Directive 2013/59/EURATOM 87

S6-160 Z. ŠimićOn the Impact to the Human Health from the Fukushima Nuclear Accident 88

Session 7 Regulatory Practice and Emergency Preparedness

S7-110 N. NovoselEU Directives in the Field of Radiological and Nuclear Safety and Their Transposition in Croatian Legislation 90

S7-119 G. PogačićCyber Security in Nuclear Power Plants – U.S. NRC Regulatory Guide 5.71 91

S7-127 Z. BazsÓImpact of TEPCO Fukushima Dai-ichi accident on severe accident management in the Slovak Republic 92

S7-128 V. Kuznetsov, Z. Drace, V. LysakovMajor Findings of IAEA/INPRO Activity on Legal and Institutional Issues for Transportable Nuclear Power Plants 93

S7-143 S. Medaković, R. BanovAssessment of NPP Krško accident impact on the population by means of the nonlinear statistical model using the RODOS software package 94


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S7-155 J. Vuković, D. Konjarek, D. GrgićRadiation doses estimation for hypothetical NPP Krško accidents without and with PCFV using RASCAL software 95

S7-164 I. Bašić, M. Kim, P. Hughes, B. K. Lim, F. D’auria, M. L. LouisIAEA Review for Gap Analysis of Safety Analysis Capability 96

S7-170 D. Konjarek, T. Bajs, D. ŠinkaPreliminary radiation doses assessment for NPP Krško SBO sequence 98

Session 8 Reactor Physics and Nuclear Fuel Cycle

S8-100 M. Matijević, D. Pevec, K. TrontlSCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations 100

S8-120 J. Haščik, G. Farkas, B. Vrban, J. Lüley, P. UrbanPower Density Determination in the VVER 440 Reactor by the Transport Codes MCNP 5 and SCALE 6 101

S8-130 M. Kromar, B. KurinčićInfluence of the Finer Radial Burnup Nodalization on the Pin Power Distribution in the PWR core 102

S8-144 M. Božič, M. Kromar, B. KurinčićFuel Reloading Strategies in a Hypothetical NPP Krško Forced Outage 103

S8-152 D. Grgić, S. Šadek, V. Benčik, D. KonjarekDecay Heat Calculation for Spent Fuel Pool Application 104

S8-154 R. Ječmenica, M. Matijević, D. Grgić

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Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs 105

S8-156 D. Zhang, F. RahnemaEfficiency and Accuracy of the Incident Flux Response Expansion Method for LaBr3 Detector Pulse Height Spectrum Calculation 106

Index of Authors 107


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Invited lectures

IL-01 S. M. ModroThe IAEA Safety Standards and SMR design features 19

IL-02 F. RomanelliThe European Roadmap to Fusion Electricity 20

IL-03 N. ČavlinaAssessment of Energy Options for CO2 Emission Reduction 21

IL-04 N. ShulyakWestinghouse Small Modular Reactor (SMR) Program 22

IL-05 N. ShulyakWestinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs 24

IL-06 Z. V. StošićReactors Project Delivery: The Value of Experience 25

IL-07 B. PetrovićProgress in Development of the I2S-LWR Concept 26

IL-08 I. SandaSafety Assessment Challenges of the New Generation Reactors – Experience from the IAEA Generic Reactor Safety Reviews 27

IL-09 J. ŠpilerNuclear Power Plant Krško 2 Project Planning and Preparation Phase 28

IL-10 C. AllisonUpdate on the SDTP sponsored Fukushima Daiichi Related Assessment Activities 29

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IL-11 T. ŽagarRadwaste Management in Small Nuclear Country – National Policy and Strategy 30

IL-12 H. ZaccaiAREVA “Sustainable Cycle Solutions” 31


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IL-01

The IAEA Safety Standards and Small and Medium Reactor Design Features

The renewed interest in nuclear energy worldwide prompted the develop-ment of small and medium size reactors (SMR). These reactors are more suitable for smaller grids as well as they promise increased safety, pos-sibly better economy, and they offer flexibility in applications (including non-power), siting and fuel cycle options. The IAEA safety standards, especially the Fundamental Safety Principles (IAEA Safety Standards se-ries No. SF-1) provide excellent principles for design of nuclear power systems, these principles however, as such, are very generic. The IAEA safety standard Safety of Nuclear Power Plants: Design (SSR-2/1) provides

specific requirements for the design of nuclear power plants, yet these requirements were developed based on best practices related to the design of current generation of reactors. On other hand, the SMRs currently under design incorporate features that are innovative but with which there is no operational experience, therefore the current standards and design requirement might not be directly applicable to the new SMR designs.This paper examines four most advanced in their development designs (NuScale, mPower, SMART, CAREM) with respect to design safety requirements of the IAEA. These designs were selected because all of them are pressurized water reactors and among SMRs they are closest relatives to the current de-signs currently being offered to the market. Therefore, the existing safety standards should be potentially applicable. Since these SMR designs are under development and most of the detailed information is not available this paper utilizes only information that is public domain. This limits the depth of the evalua-tion and mainly principal technical requirements and some general design requirements are addressed. Focus is on issues such as defence in depth, design principles (single failure criterion, fail safe design, etc.), design extension conditions, external hazards.

S. Michael ModroConsultant to the International Atomic Energy [email protected]

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IL-02

The European Roadmap to Fussion Electricity

With the reduction of CO2 emissions driving future energy policy, fusion can start market penetration around 2050 with up to 30% of electricity production by 2100. This requires an ambitious, yet realistic roadmap to-wards the demonstration of electricity production by 2050. This talk describes the main technical challenges on the path to fusion en-ergy. For all of the challenges candidate solutions have been developed and the goal of the programme is now to demonstrate that they will also work at the scale of a reactor.

The roadmap has been developed within a goal-oriented approach articulated in eight different Mis-sions. For each Mission the critical aspects for reactor application, the risks and risk mitigation strate-gies, the level of readiness now and after ITER and the gaps in the programme have been examined with involvement of experts from the ITER International Organization, Fusion for Energy, EFDA Close Sup-port Unites and EFDA Associates. High-level work packages for the roadmap implementation have been prepared and the resources evaluated. ITER is the key facility in the roadmap and its success represents the most important overarching objec-tives of the EU programme. A demonstration fusion power plant (DEMO), producing net electricity for the grid at the level of a few hundreds MW is foreseen to start operation in the early 2040s. Following ITER, it will be the single step to a commercial fusion power plant. Industry must be involved early in the DEMO definition and design. The evolution of the programme requires that industry progressively shifts its role from that of provider of high-tech components to that of driver of the fusion development. Industry must be able to take full responsibility for the commercial fusion power plant after successful DEMO operation. For this reason, DEMO cannot be defined and designed by research laboratories alone, but requires the full involvement of industry in all technological and systems aspects of the design. Europe should seek all the opportunities for international collaborations. Some of the ITER parties have a very aggressive programme in fusion and Europe can clearly benefit by the participation in the design, construction and operation of their facilities. Already the Broader Approach with Japan is a good example of a positive collaboration that can give further advantages on the time scale considered here.The talk will also address the needs in the area of education and training and basic research.

Francesco RomanelliEFDA & EFDA-JET, CSUCulham Science Centre, Oxford, OX14 3DB, United [email protected]


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IL-03

Assesment of Energy Options for CO2 Emission Reduction

Since the 1992 Earth Summit in Rio de Janeiro, global anthropogenic CO2 emissions grew by 52% which caused an increase in of 10.8% in the CO2 concentration in the atmosphere, and it tipped the 400 ppm mark in May 2013 . The Fifth Assessment Report on climate impacts from the Intergov-ernmental Panel on Climate Change (IPCC) confirmed earlier warnings that climate change is already stressing human communities, agriculture, and natural ecosystems, and the effects are likely to increase in the future. While European Union has long been committed to lowering carbon

emissions, this places additional pressure on current EU goals for energy sector that include significant reduction of CO2 emissions. Current EU commitment has been formalized in so-called “20-20-20” plan, reducing carbon emissions, increasing energy efficiency and increasing energy production from renewa-bles by 20% by 2020. Some EU member states are even more ambitious, like United Kingdom, planning to reduce carbon emissions by 80% by 2050. Bulk of carbon reduction will have to be achieved in energy sector.In the power industry, most popular solution is use of solar and wind power. Since their production var-ies significantly during the day, for the purpose of base-load production they can be paired with gas-fired power plant. Other possible CO2-free solution is nuclear power plant. In this invited lecture, predicted cost of energy production for newly built nuclear power plant and newly built combination of wind or solar and gas-fired power plant are compared. Comparison was done using Levelized Unit of Energy Cost (LUEC). Calculations were performed using the Monte Carlo method. For input parameters that have biggest uncertainty (gas cost, CO2 emission fee) those uncertainties were addressed not only through probability distribution around predicted value, but also through different scenarios.

Nikola ČavlinaUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected]

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IL-04

Westinghouse Small Modular Reactor (SMR) Program

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) in which all primary components associated with the nuclear steam supply system, including the steam generator and the pressurizer, are housed within the reactor vessel. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000™ plant design with a compact containment

that houses the integral reactor vessel and the passive safety systems. This paper describes the design and functionality of the Westinghouse SMR, the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR.Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design drivers include safety, economics, reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements.The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000™ reactor, and provides mitigation of all design basis accidents without the need for offsite AC electrical power for a period of seven days. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reac-tors (LWR) that utilized the economies of scale to reach economic competiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is im-plemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology.The integral Westinghouse SMR design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. This compact containment will be completely submerged in water during power operation providing a heat sink for postulated accidents which also aides the heat removal and provides an additional radionuclide filter. For protection against external threats, the containment vessel and plant safety systems are located below ground level. At a diameter of 32 feet, approximately 25 of the Westinghouse SMR containment vessels can fit within the envelope of the AP1000™ containment building.The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and will be fueled by a derivative of the successful

Nick ShulyakWestinghouse Electric [email protected]


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17x17 Robust Fuel Assembly (RFA) product. An 89 assembly core with an active height of 2,4 m (8 feet) will provide a 24 month operating cycle with a power output of 800 MWt. Derived from the AP1000™plant and adapted to operate inside the reactor pressure vessel, 37 control rod drive mechanisms provide reac-tor shutdown and reactivity control capabilities. Eight seal less pumps provide a nominal reactor coolant flow of 100,000 gallons per minute. An innovative evolution of a straight tube steam generator produces a saturated mixture that is delivered to a steam separating drum located outside of the containment ves-sel. The steam generator along with the integral pressurizer is attached to the reactor vessel with a single closure flange located near the center of gravity of the reactor assembly and is designed to be removed during refueling operations. Like the AP1000™ plant, the Westinghouse SMR relies on the natural forces of gravity and natural circulation to provide core and containment cooling during accident conditions.At approximately one fifth the net electrical output of the AP1000™ plant, the Westinghouse SMR is de-signed to address infrastructure challenges associated with replacing aging fossil fuel plants by providing a safe, clean and reliable energy source. The challenges associated with economies of scale are offset with a compact and simplified plant design, rail shippable components and modular construction.

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IL 5

Westinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs

Westinghouse Electric Company once again sets a new industry standard with the AP1000® reactor. Historically, Westinghouse plant designs and technology have forged the cutting edge of worldwide nuclear technology. Today, about 50 percent of the world’s 440 nuclear plants are based on Westinghouse technology. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace,

and is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). The AP1000 features proven technology, innovative passive safety systems and offers:

• Unequalled safety• Economic competitiveness• Improved and more efficient operations

The AP1000 builds and improves upon the established technology of major components used in current Westinghouse-designed plants with proven, reliable operating experience over the past 50 years. These components include:

• Steam generators• Digital instrumentation and controls• Fuel• Pressurizers• Reactor vessels

Simplification was a major design objective for the AP1000. The simplified plant design includes overall safety systems, normal operating systems, the control room, construction techniques, and instrumentation and control systems. The result is a plant that is easier and less expensive to build, operate and maintain.The AP1000 design saves money and time with an accelerated construction time period of approximately 36 months, from the pouring of first concrete to the loading of fuel. Also, the innovative AP1000 features:

• 50% fewer safety-related valves• 80% less safety-related piping• 85% less control cable• 35% fewer pumps• 45% less seismic building volume

Eight AP1000 units under construction worldwide – Four units in China – Four units in the United States

Nick ShulyakWestinghouse Electric [email protected]


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IL-06

Reactors Project Delivery: The Value of Experience

State of Affairs• Energy Potential and Density versus Environmental Load of differ-

ent Energy Sources• Development of Fuel into Energy/Electricity Generation• Production Costs of Electricity• Contributions of Nuclear Energy to Security of Energy Supply

• Recent Nuclear Development• Public Support growing again

Projects Status• Reactors under Construction• Different Projects Industrial Schemes• Projects Overview

The Value of Experience• Licensing• Standardization on Early Engineering Activities• Supply Chain and Manufacturing of Heavy Components• Installation• Procurement

Zoran V. StošićAREVAKoldenstrasse 16, 91052 Erlangen, [email protected]

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IL-07

Progress in Development of I2S-LWR Concept

The paper will present the progress in developing the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept. This new concept aims to combine the competitive economics of a large nuclear power plant, with enhanced safety achieved by the integral primary circuit configuration (previously considered only for PWRs with power levels not exceeding several hundred MWe), and with enhanced accident tolerance (to address concerns after the Fukushima Dai-Ichi accidents). Several new technolo-gies are being developed to enable this concept, including novel silicide fuel and micro-channel primary heat exchangers. This project is per-

formed by a multi-disciplinary multi-organization team led by Georgia Tech, including academia, a na-tional laboratory, nuclear industry, and a power utility, wit expected participation of the University of Zagreb.

Keywords: PWR, integral reactor vessel, inherent safety, economics, enhanced accident tolerance

Bojan PetrovićNuclear and Radiological Engineering, Georgia Institute of Technology770 State St., Atlanta, GA 30332-0745, [email protected]


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IL-08

Safety Assesment Challanges of the New Generation Reactors – Experience from the IAEA Generic Reactor Safety Reviews

New reactor designs that are available, or are expected to be available in short term on the market, have been developed to minimize potential risk to the public at overall competitive cost. New design features have also generated new challenges on the verification of the safety analysis meth-odologies used for the design and licensing. In the recent years, as a part of the design review services, IAEA has provided services of the early evalu-ation of a vendor’s submission safety case of a new nuclear power plant against the IAEA Safety Standards at the fundamentals and requirements

level – Generic Reactor Safety Reviews (GRSR). In all of the reviewed new reactor designs, combination of the PSA and deterministic methodologies have been used for the development of the safety cases for the new reactors, combining conservative and best estimate methodologies. Harmonized approach of the group of the international experts and the IAEA revealed some issues during GRSR regarding impact of the new design features on the safety analysis presented as the part of the safety case. Overview of the most interesting issues shall be presented at the CRA forum.

Irina SandaConsultant to the International Atomic Energy [email protected]

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IL-09

Nuclear Power Plant Krško 2 Project Planning and Preparation Phase

GEN energija is planning a new unit of Krško NPP. Feasibility and ac-ceptability studies showed that the project would be the most appropri-ate sustainable solution for the future Slovenian electrical energy system. However, there is no strategic decision on the project yet.In Slovenia, the situation in electricity supply has intensified in recent years, electricity consumption has increased, and the situation of the country has changed due to the global economic crisis. When the econo-

my recovers, electricity consumption will begin to climb again. Today Slovenia is a net importer of elec-tricity, but with the addition of one or two nuclear units, this could change, as there is significant demand in the surrounding countries, in particular Italy.In a new draft of the Development Strategy of the Republic of Slovenia from 2013 to 2020, nuclear energy is recognized as a sustainable energy source and a key technology to low carbon society. This position means a basis for current as well as extended operation of the existing NPP Krško. Furthermore, it also gives a basis for the new unit of Krško NPP. Nevertheless, the new unit would still need to be proposed and confirmed in the forthcoming Energy Concept of Slovenia. The process of preparing the new Energy Concept of Slovenia will start at the beginning of next year and it will presumably take a year for the Gov-ernment to issue a new long term energy strategy.

Keywords: nuclear power, energy strategy, feasibility studies, sustainability

Jože ŠpilerGEN energija, d.o.o Vrbina 17, 8270 Krško, [email protected]


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IL-10

Update on the SDTP Sponsored Fukushima Daiichi Related Assesment Activities

Following the accident at Fukushima Daiichi, Innovative Systems Soft-ware and several other members of an international software develop-ment and training group (SDTP) started an assessment of the possible core/vessel damage states of Units 1-3. This assessment, using a reference RELAP/SCDAPSIM Laguna Verde BWR model developed by CNSNS, the Mexican nuclear regulatory body, was presented initially to the IAEA emergency response team for Fukushima in March of 2011. Our assess-ment for the IAEA indicated that significant fuel melting, fuel slumping, and lower head failure was likely for Units 1 and 3. The results for Unit 2

were inconclusive because of the complex thermal hydraulic conditions at the time of likely fuel melting. Since that time the SDTP related assessment activities have continued on three main fronts: (a) continued analysis using our representative Laguna Verde model to determine the likely failure modes leading to an un-intentional depressurization of the vessel during a SBO in a BWR, (b) development of improved RELAP/SCDAPSIM models to treat the likely mode of lower core support structure melting and failure, and (c) design studies for proposed fuel melting and relocation experiments in Japan to support model development and cleanup related activities. The presentation gives a brief summary and discussion of these activities.

Chris AllisonInnovative Systems Software, LLC 3585 BriarCreek Lane, Ammon Idaho, [email protected]

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IL-11

Radwaste Management in Small Nuclear Country – National Policy and Strategy

The lecture will briefly present the Slovene nuclear program and its le-gal framework focused on the radioactive waste management policy and strategy aspect. Slovenia is an example of small EU member state with small shared nuclear power program demonstrating safe, secure and ef-ficient management of radioactive waste. Different principles of radioactive waste management will be discoursed; among others including: minimization of waste generation, the polluter

pays principle, safe storage followed by final disposal and also new findings on research and development of storage, disposal and recycling of radioactive waste.

Tomaž ŽagarARAOCelovška cesta 182, 1000 Ljubljana, [email protected]


31


IL-12

AREVA „Sustainable Cycle Solutions“

Almost three years after the Fukushima events, the nuclear industry has shown its capacity to respond to first additional safety requirements and continues working closely with the regulators and involved stakeholders for further improvements. Continued growth is still anticipated and sev-eral international organizations have confirmed 2035 scenarios close to or above 600 GWe of installed nuclear capacity. Securing and optimizing the fuel cycle is crucial to ensure the sustainability of such a nuclear pro-gram in the medium term. Rethinking fuel cycle schemes is also decisive

to prepare long term transitions to next generations of reactors. R&D and Innovation shall remain a cor-nerstone of this sustainable development.Countries with significant or growing nuclear fleets have confirmed their commitment or plans to recy-cle. This is notably the case in China, France, India, Japan, the Netherlands, South Korea, and the Russian federation. Most of other nuclear countries are currently rethinking their used fuel management, taking into account;

• the geopolitical context and the international framework, including multilateral agreements,• the legacy of their past fuel cycle policies, • the cornerstones of energy policies’ evolution such as the targets set for energy independence, as

well as the level of their ambition in terms of nuclear park evolution, • the involvement of concerned stakeholders in the decision making process,• the regulatory and legal frameworks evolutions,• the operational challenges that the nuclear actors are currently facing, in particular the increasing

constraints related to reactor pool storage, or to very long term dry storage.

With the foreseen nuclear development, the used fuel backlog will continue to grow for several decades. Bolstered by a 40 years experience, recycling has been contributing to further increase safety while offer-ing main operational, environmental, global acceptance benefits, among others . However even though recycling capabilities are on track to expand in some countries there will be a growing need, to move forward efficiently, to smartly mix proven and evolving solutions (recycling, on site dry storage, pools, centralized storage, advanced technologies). These shall be combined in an optimized manner taking into account key criteria related to non proliferation, minimization of environmental impact, economics, fleet performance, responsibility towards future generations… The recommendations from Safety Authorities to deepen the question of very long term dry storage have also reactivated the debate about the available

Henri ZaccaiAREVATour AREVA, 1 Place Jean-Millier, F-92084 ParisLa Defense [email protected]

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options for the short and long terms, including recycling, which is therefore likely to play an increasing role, especially if flexibility is required to ensure a cost-effective fast reactor ramp-up. Such an approach is aimed at meeting needs that may differ, even diverge depending on the concerned stakeholders.The paper will in particular illustrate how AREVA “Sustainable Cycle Solutions” may optimize fuel cycle schemes for the short and longer terms and fulfill the stakeholders’ needs. It will also pinpoint the related limits and conditions that have to be met to fulfill the above mentioned criteria.


33


Session 1 Energy Planning and Nuclear Option

S1-148 V. Knapp, D. PevecA View on the Future of Nuclear Fission Energy 34

S1-149 V. Knapp, D. Pevec, M. Matijević, D. LaleMolten Salt Thorium Reactor – A Promising Nuclear Technology to Stop Global Warming 35

S1-162 Ž. TomšićFrom Market Uncertainty to Policy Uncertainty for Investment in Power Generation: Real Options for NPP on Electricity Market 36

Session 1: Energy Planning and Nuclear Option (EPNO)34

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S1-148

A View on the Future of Nuclear Fission Energy

Recent publications discussing the role of nuclear energy in contributing to carbon emission reduction take different approaches and reach very different conclusions. For the pessimistic approach the nuclear contri-bution in the year 2050 would be on the unimportant level of 8 EJ/year, whilst the optimistic approach with early introduction of fast breeders sees the nuclear contribution in 2060 on the level 144 EJ/year with mas-sive build-up of breeder reactors in the years 2030-2060, reaching the nu-clear capacity of 5372 GW in the year 2060. We do not find the optimis-tic strategy acceptable from political, safety and technical grounds. We show instead that a technologically more conservative nuclear build-up

in the years 2025-2065 with proven conventional reactors using once through fuel cycle without fuel re-processing could reach 3300 GW on the uranium resources as known in 2008. With this concept nuclear contribution of 94 EJ/year would be reached by 2065, many times more than the pessimistic estimate, while integral CO2 emission savings would be about 500 GtCO2. This shows that essential nuclear con-tributions is possible without the use of plutonium and fast breeders, technology not ready for climate-critical next 50 years and not acceptable in present political environment.

Keywords: nuclear fission energy, nuclear energy strategy, carbon emission savings

Vladimir Knapp, Dubravko PevecUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected], [email protected]

Session 1: Energy Planning and Nuclear Option (EPNO) 35

10TH INTERNATIONAL CONFERENCE: NUCLEAR OPTION IN COUNTRIES WITH SMALL AND MEDIUM ELECTRICITY GRIDS1 – 4 June 2014, Zadar, C

roatia

S1-149

Molten Salt Thorium Reactor – A Promising Nuclear Technology to Stop Global Warming

If we take seriously the repeated demands by IPCC for timely contribu-tion to required reduction of carbon emission we have only few decades to act. Important quantitative results are given in the study of Meinshausen et al. In order to keep global temperature rise till the end of century be-low 2 ºC, with the probability not exceeding this limit of 25% (50%) , the world would have to limit cumulative emission in the interval 2000-2050 to 1000 Gt (1440 Gt). As the emission of CO2 during the interval 2000-2006 amounted to 234 Gt of CO2 we can appreciate the magnitude of the problem. Assuming continuation of average annual emission of 36.3 Gt we would exhaust emission budget by 2027 respectively by 2039 (for 50% limit).Other studies by leading climatologists are supporting time available for effective action or even shortening it. In our recent paper we explored a potential of the proven light water reactors without fuel reprocessing and plutonium recycle to contribute essentially to reduction of carbon emis-sion. We selected developed and established nuclear technology because it is available now, in spite of the economic and conceptual limitations. Thus we built on the results of 2010 study determining the maximum nu-clear contribution possible with light water reactors of Generation 3+ as-

suming complete consumption of uranium reserves as estimated in 2008 in the years 2025-2065. It turns out that by 2065 that strategy can give a very significant nuclear carbon free share of about 39% in the projected 2065 “business as usual” energy consumption. We discussed very briefly the question how to proceed after 2065. One possibility would be sodium cooled Fast Breeder Reactors (FBR) to be launched with plutonium accumulated in the operation of LWR reactors until and after 2056. In our judgement advantage would have Molten Salt Thorium Reactors (MSTR), one of Generation 4 selection, having outstanding research and development attention. Early development, dated from the years 1964-69, with their superior safety properties and also the fact elaborated below that they could be introduced earlier than possible long term alternative FBR reactors. MSTR reactors could be introduced on the large scale about 2045, due to smaller fissile launching requirement. A transition to thorium fuel after 2045 would remove fuel limitation to practically any foreseeable time.

Keywords: molten salt reactor, thorium, nuclear energy strategies

Vladimir Knapp, Dubravko Pevec, Mario MatijevićUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected], [email protected], [email protected]

Dinka LaleUniversity of Dubrovnik, Department of Electrical Engineering and ComputingĆira Carića 4, 20000 Dubrovnik, [email protected]

Session 1: Energy Planning and Nuclear Option (EPNO)36

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S1-162

From Market Uncertainty to Policy Uncertainty for Investment in Power Generation: Real Options for NPP on Electricity Market

In the electricity sector, market participants must make decisions about capacity choice in a situation of radical uncertainty about future market conditions. Sector is normally characterised by non-storability and peri-odic and stochastic demand fluctuations. In these cases capacity deter-mination is a decision for the long term, whereas production is adjusted in the short run. Capacities need to be installed well in advance (decision for investment even earlier because of long construction time and even longer in case of NPP to prepare all needed legal, financial and physical

infrastructure), at times when firms face considerable demand and cost uncertainty when choosing their capacity. Paper looks on the main contributions in investment planning under uncertainty, in particular in the electricity market for capital intensive investments like NPP. The relationship between market and non-market factors (recent UK policy example) in determining investment signals in competitive elec-tricity markets was analysed. Paper analyse the ability of competitive electricity markets to deliver the desired quantity and type of generation capacity and also investigates the variety of market imperfections operating in electricity generation and their impact on long-term dynamics for generation capacity, the most capital-intensive of the liberalised functions in the electricity supply industry. Paper analyses how price formation influences investment signals. Today, investment decisions are made by several opera-tors that act independently. Number of factors (including market power, wholesale price volatility, lack of liquidity in the wholesale and financial market, policy and regulatory risks etc.) contribute to polluting the price signal and generating sub-optimal behaviour.Climate change policies can easily distort market signals, insulating renewables generation from market dynamics. This in turn reduces the proportion of the market that is effectively opened to competitive forces. When renewable support policies are undertaken, investments in conventional technologies suf-fer (especially in capital intensive investment like NPP). This produces distorting effects on the genera-tion mix. Policy intervention, rather than market forces, is able to select artificially winners and losers, thus potentially undermining, in the long run, the necessary diversity of the energy mix.

Keywords: power generation, nuclear power plant, electricity market, market uncertainty, investments, generation capacity

Željko TomšićUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected]


37


Session 2 Power Reactors and Technologies

S2-106 M. Kim, S. Hong, Y. Kim, D. Shin, J. Lee, G. ParkStatus of Thermo-Fluid Experimental Research on VHTR in Korea 38

S2-116 H. Grganić, B. GlaserComparison of AP1000 and Generation II PWR with an Accent on Safety Systems and Coping with Station Blackout 39

S2-141 B. Niceno, M. PouchonThermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides 40

S2-151 D. Grgić, T., Fancev, V. Benčik, S. ŠadekHot Leg Streaming Calculation for Two-Loop PWR Plant 41

S2-216 A. Manera, M. J. MemmottDesign and Trade-off Studies of the Passive Decay Heat Removal System (DHRS) of the Integral Inherently Safe LWR (I2S-LWR) 42

Session 2: Power Reactors and Technologies (PRT)38

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S2-106

Status of Thermo-Fluid Experimental Research on VHTR in Korea

For reasons related to licensing uncertainty, economic slowdown, and questionable financial backing, no new nuclear facility projects have been undertaken in the United States since the Three Mile Island Incident in 1979; however, a need for such facilities (both nuclear power plants and nuclear fuel facilities) continues and various incentives leading to the start of a nuclear renaissance have occurred. One incentive is a complete over-haul by the US Nuclear Regulatory Commission of the earlier two step licensing process under 10 CFR 50. The earlier approach required first a construction permit and then an operating license, whereas the new ap-proach allows a more streamlined (one step) combined license (COL) ap-proach utilizing Standard Design Certifications via the regulatory frame-work created by 10 CFR 52. Other incentives include US Government backed loan guarantees as well as private company contributions.One aspect to the new process has been consideration and implementa-tion of many new topic-specific regulations and industry standards which have continued to evolve during the past 30 years in spite of the lack of new plant design and construction activity. Therefore, an Owner attempt-ing a new nuclear facility project under 10 CFR 52 needs to address a myriad of new requirements previously unconsidered.Several new projects including both power plants and fuel facilities have begun the new licensing process with its many new requirements to con-sider, but a uranium enrichment facility has run the gamut first. This paper will summarize many of the lessons learned from designing, con-structing and testing this first new nuclear facility to be built in the US in over 30 years.

Keywords: new construction, lessons learned, 10 CFR 50, 10 CFR 52, com-bined license, nuclear renaissance

Min-Hwan Kim, Seong-Deok Hong, Yong-Wan KimKorea Atomic Energy Research Institute989-111 Daedeok-daero, Yuseong-gu, Daejeon, Republic of [email protected], [email protected], [email protected]

Dong-Ho Shin, Jeong-Hun LeeSeoul National University, Department of Nuclear Engineering1 Gwanak-ro, Gwanak-gu, Seoul, Rep. of [email protected], [email protected]

Goon-Cherl ParkKEPCO International Nuclear Graduate School658-91 Haemaji-ro, Seosaeng-myeon, Ulju-gun, Usan, Rep. of [email protected]

Session 2: Power Reactors and Technologies (PRT) 39


roatia

S2-116

Comparison of AP1000 and Generation II PWR with an Accent on Safety Systems and Coping with Station Blackout

Generation III reactors are advanced successors of the Generation II nu-clear reactors, incorporating evolutionary improvements in fuel technol-ogy, thermal efficiency, passive safety, and standardized design.This paper gives comparison of two Westinghouse’s designs – Generation III PWR AP 1000 and Generation II PWR Krško NPP. Comparison is fo-cused on safety systems of both reactors.

In the case of Station Blackout AP1000 can achieve and maintain safe shutdown for 72 without any opera-tor actions. Timelines for Station Blackout events are compared and discussed in detail. Furthermore, this paper gives response comparison of different systems and structures (Decay Heat Removal System, Spent Fuel Pit) following the Station Blackout for both AP1000 and Krško NPP.Currently is NEK implementing Safety Upgrade Program (PNV – Program nadogradnje varnosti), which will enhance nuclear safety through several modifications that are already implemented (e.g. third diesel generator) or will be implemented in the next years. Influence of those modifications is also discussed and compared with AP1000.

Keywords: Krško NPP, AP1000, safety systems, passive safety, Station Blackout, Safety Upgrade Program

Hrvoje Grganić Bruno GlaserNuclear Power Plant KrškoVrbina 12, 8270 Krško, [email protected]


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S2-141

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

The accident in Fukushima has drastically shown the drawbacks of Zir-caloy claddings despite their beneficial properties in normal use. The ef-fect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an al-ternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel

cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on ther-mal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident manage-ment guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these is-sues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from fundamental point of view, to prove why SiC as the material influences it. Finally, proposed research direction for further studies on thermal-hydraulic considerations of SiC material for cladding are outlined.

Keywords: nuclear fuel cladding, hydrogen production, silicon carbide, zircaloy, thermal-hydraulics

Bojan Niceno, Manuel PouchonPaul Scherrer Institute, Nuclear Energy and Safety Department5232 Villigen PSI, [email protected], [email protected]

Session 2: Power Reactors and Technologies (PRT) 41


roatia

S2-151

Hot Leg Streaming Calculation for Two-Loop PWR Plant

Thermal stratification in hot leg pipe (often called hot leg streaming (HLS)) of PWR plant is caused by different heating of core fluid streams in fuel assemblies having different power production. Fluid streams hav-ing different temperature at core exit, after some mixing, enter hot leg pipe. Any perpendicular plane downstream hot leg pipe will have some kind of 2D temperature distribution and temperature measured by im-mersed RTD will depend on circumferential position and insertion depth of RTD detector. In order to get representative average temperature of the water typically 3 different circumferential measurement positions were used. The model for HLS calculation is developed based on core exit tem-peratures calculated by PARCS code and ANSYS-FLUENT CFD model of outlet plenum and hot legs. The scooping calculation was performed to determine HLS temperature profiles for NPP Krško.

Keywords: hot leg streaming, PARCS, FLUENT, temperature profiles

Davor Grgić, Tomislav Fancev, Vesna Benčik, Siniša ŠadekUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected], [email protected], [email protected], [email protected]


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S2-216

Design and Trade-off Studies of the Passive Decay Heat Removal System (DHRS) of the Integral Inherently Safe LWR (I2S-LWR)

The Integral, Inherently Safe Light Water Reactor (I2S-LWR) is a novel pressurized water reactor (PWR) concept being developed by a multi-in-stitutional team led by Georgia Tech, under the Department of Energy’s Nuclear Energy University Programs Integrated Research Projects (DOE NEUP IRP). The I2S-LWR aims at delivering an electric power output level comparable to that of large LWRs (approximately 1,000 MWe), while at the same time achieving an overall level of safety that is enhanced with respect to large Generation III+ LWRs. In order to accommodate this, the I2S-LWR incorporates several innovative safety features, including an integral reactor pressure vessel and enhanced passive decay heat removal systems. In particular, these passive decay heat removal systems focus on removal of decay heat during potential transient scenarios without the need for operator action, offsite water, or AC power. The target for these passive safety systems is to provide decay heat removal capabilities for an indefinite period of time for a large subset of accident scenarios, while

demonstrating an enhanced coping time for the most severe accident scenarios relative to passive LWR technology currently in use ( >7 days). The primary decay heat removal system (DHRS) for the I2S utilizes air as the ultimate heat sink, with two water loops which circulate under natural circulation to transfer decay heat to the ultimate heat sink. Heat transfer and water flow is initialized by the opening of a fail open valve in the primary loop, and heat is then transferred between loops via helical coil exchangers, and ultimately released to the atmosphere through an air/water exchanger. This paper discusses the base de-sign of the DHRS, including potential design enhancements that can be included in order to improve heat transfer performance. Additionally, this paper presents a series of sensitivity studies that were conducted to establish the feasibility and operational space of the DHRS and to identify potential design limitations. A RELAP5 model was created to further verify the performance of the base design.

Annalisa ManeraDepartment of Nuclear Engineering & Radiological Sciences, University of Michigan2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104, [email protected]

Matthew J. MemmottWestinghouse electric Company1000 Westinghouse Drive, Cranberry Township, PA 16066, [email protected]


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Session 3 Nuclear Safety

S3- 101 A. ProšekSensitivity analysis of LOFT L2-5 test calculations 45

S3- 117 G. Vayssier, I. BašićSystematic Review of Accident Management Programs – Principles, Experiences 46

S3-121 B. MarinovaSeismic margin assessment for nuclear facilities of Kozloduy NPP 47

S3-125 I. Horvatovic, C. Batra, M. Cherubini, A. Petruzzi, F. D’auria, T. Bajs Benchmark analysis of EBR–II shutdown heat removal test SHRT–17 48

S3- 135 M. Uršič, M. Leskovar, R. MeignenAnalysis of Effect of Sodium Thermo-Dynamical Properties in Fuel-Coolant Interaction 49

S3- 137 M. Mihalina, S. Špalj, B. Glaser, R. Jalovec, G. JankovićNPP Krško SAMG Upgrade 50

S3- 139 S. Šadek, D. GrgićContainment Modelling with the ASTEC Code 51

S3- 146 K. ManchevaMain Aspects and Results of Level 2 PSA for KNPP WWER-1000/B320 52

S3- 150 D. Grgić, T. Fancev, S. ŠadekCalculation of hydrogen concentration in containment during LOCA accident 53

S3- 153 V. Benčik, D. Grgić, N. Čavlina, S. ŠadekOptimization of OPDT Protection for Overcooling Accidents 54

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S3- 157 R. Banov, Z. Šimić, D. ŠtercShort overview of PSA quantification methods, pitfalls on the road from approximate to exact results 55

S3- 163 D. Slovenc, I. Bašić, L. ŠtrubeljSimulation and validation of a full and mini flow surveillance test in NEK with best estimate HPIS model in APROS 56

Session 3: Nuclear Safety (NS) 45


roatia

S3-101

Sensitivity analysis of LOFT L2-5 test calculations

The uncertainty quantification of best-estimate code predictions is typi-cally accompanied by a sensitivity analysis, in which the influence of the individual contributors to uncertainty is determined. The objective of this study is to demonstrate the improved fast Fourier transform based method by signal mirroring (FFTBM-SM) for the sensitivity analysis. The sensitivity study was performed for the LOFT L2-5 test, which simulates the large break loss of coolant accident. There were 14 participants in the

BEMUSE (Best Estimate Methods – Uncertainty and Sensitivity Evaluation) programme, each perform-ing a reference calculation and 15 sensitivity runs of the LOFT L2-5 test. The important input parameters varied were break area, gap conductivity, fuel conductivity, decay power etc. For the influence of input parameters on the calculated results the FFTBM-SM was used. The only difference between FFTBM-SM and original FFTBM is that in the FFTBM-SM the signals are symmetrized to eliminate the edge effect (the so called edge is the difference between the first and last data point of one period of the signal) in cal-culating average amplitude. It is very important to eliminate unphysical contribution to the average am-plitude, which is used as a figure of merit for input parameter influence on output parameters. The idea is to use reference calculation as ‘experimental signal’, ‘sensitivity run’ as ‘calculated signal’, and average amplitude as figure of merit for sensitivity instead for code accuracy. The larger is the average amplitude the larger is the influence of varied input parameter.The results show that with FFTBM-SM the analyst can get good picture of the contribution of the pa-rameter variation to the results. They show when the input parameters are influential and how big is this influence. FFTBM-SM could be also used to quantify the influence of several parameter variations on the results. However, the influential parameters could not be identified nor the direction of influence. The results suggest that FFTBM is especially appropriate for a quick quantitative sensitivity analysis in which several calculations and/or influence of sensitive parameters need to be compared.

Keywords: sensitivity analysis, FFTBM-SM, LOFT L2-5, thermal-hydraulic calculation, signal mirroring

Andrej ProšekJožef Stefan InstituteJamova cesta 39, SI-1000 Ljubljana, [email protected]

Session 3: Nuclear Safety (NS)46

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S3-117

Systematic Review of Accident Management Programs – Principles, Experiences

Although all plants have some form of accident management, there is not always a proper review of the accident management program neither of its products, i.e. the various procedures and guidelines. Moreover, such reviews are often limited to Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG). More complex events, which include large damage on the site, require additional tools and pro-cedures / guidelines.The present paper describes a new review method that covers this larger area and is capable to identify problems and shortcomings, and offers so-lutions for those. It basically exists of a three-tier approach:

1. interviews with the national regulator and/or the plant to evaluatethe scope of the accident management as required by the national regulation and in comparison with international regulation;

2. interviews with the plant staff to discuss the technical basis of the accident management program and its implementation; and

3. observation of an exercise to test the capability of the plant staff to execute the accident manage-ment procedures and guidelines, as well as the value of the exercise for such test.

The method is an extension of the IAEA ´Review of Accident Management Program`, which is limited to review of EOPs and SAMG. It is based on extensive experience with plant reviews.

Keywords: accident management, review method, principles, experiences

George VayssierNSC Netherlands, Kamperweg 1, 4417PC Hansweert, The [email protected]

Ivica BašićAPoS d.o.o.Repovec 23 b, 49210 Zabok, [email protected]



roatia

S3-121

Seismic margin assessment for nuclear facilities of Kozloduy NPP

In accordance with the decision of the European Commission and EN-SREG Declaration of 13 May 2011, all nuclear power plants in the Euro-pean Union were subjected to a stress test. The stress test is defined as a targeted reassessment of the safety margins of nuclear power plants in the light of the events which occurred at Fukushima: extreme natural events challenging the plant safety functions and leading to a severe accident.Seismic margins assessment is based on the analysis of the seismic resist-

ance of the equipment, which is important for safety and participates in mitigation of accident scenarios. Seismic margin is determined on the basis of the prescribed limits of seismic accelerations that any nu-clear facility can withstand without severe fuel damage and radioactive release into the environment.The determining of the weak points and boundary effects in case of seismic action is done based on the data from the seismic PSA Level 1. Based on the calculated median values of the probable seismic devia-tions, the ranges of probable seismic action have been determined for which the resistance of the different nuclear facilities is assessed.The safety margins re-assessment should define the nuclear facility ultimate capacity, i.e. to determine the values of accelerations for which the SSC failures would result in non-availability of the safety functions, and fuel damage would be inevitable. The assessment of this acceleration value is done using the data of the seismic analyses as performed on design stage, accounting for the dynamic response and actual spatial dimensions of the civil structures and for the materials properties.For the purpose of the KNPP nuclear facilities safety re-assessment, the seismic capacity is accepted to be determine by the value of seismic acceleration, for which it can be ascertained with 95% certainty that the safety factor obtained at the respective seismic acceleration is not lower than 1.The purpose of this report is to present the approach, main results and conclusions of the seismic margin assessment for „Kozloduy“ NPP.

Keywords: seismic margin, PSA, stress tests

Bozhana MarinovaRisk Engineering Ltd.10 Vihren Str., 1618 Sofia, Bulgaria [email protected]


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S3-125

Benchmark analysis of EBR–II shutdown heat removal test SHRT–17

Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The aim of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site. In the frame of this project, benchmark analysis of one of the EBR-II shut-down heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed at the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) in Pisa, Italy. The aim of this paper is to present modelling of EBR-II reactor design using RELAP-3D, and to present results of the transient analysis of SHRT-17. Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX). After achievement of ac-ceptable steady-state results, transient analysis was performed. Start-ing from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was

scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary cool-ant pump that normally had an emergency battery power supply was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable leveles and cooled the reactor down safely at decay heat power levels.Thermal-hydraulics characteristics and plant behaviour was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL. The plan is to improve the model in future with replacing current models of core and pool with 3D mod-els, and, eventually, coupling with neutronic codes for more accurate results.

Keywords: EBR-II, Sodium-cooled fast reactors, SHRT-17, RELAP-3D, protected loss-of-flow

Ivan Horvatovic, Chirayu Batra, Marco Cherubini, Alessandro Petruzzi, Francesco D’AuriaUniversity of Pisa,Gruppo di Ricerca Nucleare San Piero a GradoVia Livornese 1291, Pisa, [email protected], [email protected], [email protected], [email protected], [email protected]

Tomislav BajsEnconet d.o.o.Miramarska 20, Zagreb, [email protected]



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S3-135

Analysis of Effect of Sodium Thermo-Dynamical Properties in Fuel-Coolant Interaction

The Generation IV International Forum has selected six technologies for future nuclear reactors, including sodium cooled fast reactor. The con-struction of several demonstration-scale sodium cooled fast reactors is planned during the 2020s in different countries.In the frame of safety studies for the demonstration-scale reactors, it is important to estimate the risk for the environment in case of a severe ac-cident. An unprotected transient over power or a loss of coolant flow may result in core melt. A vapour explosion may occur during a core melt ac-cident, when the rapid and intense heat transfer follows the interaction between the molten corium and the sodium. Potentially severe dynamic loadings on surrounding systems, structures and components could be induced.In experiments an important effect of the sodium sub-cooling on the be-haviour of the melt-sodium interaction was identified. The vapour explo-sion probability and efficiency for higher sub-cooling was lower than for

lower sub-cooling. The physical properties of sodium, which strongly affect the melt-sodium heat trans-fer, and the melt solidification process, which strongly affects the energy efficiency during the explosion, are identified as the reason for the observed behaviour.The analytical research of the melt-sodium interaction with fuel-coolant interaction codes requires ad-equate modelling of the sodium thermo-dynamic and transport properties, the melt-sodium heat transfer and the effect of melt solidification. One of the computer codes, which have the potential to simulate vapour explosions with sodium, is the MC3D code. The MC3D code is being developed and managed by IRSN, France, with support of CEA, France, and EDF, France. Since 2005 also our Reactor Engineering Division, Jožef Stefan Institute, is participating in the validation and the development of the code.The aim of the paper is to analyse the tables for the sodium thermo-dynamic and transport properties used in the MC3D code. Based on the sodium physical properties and current understanding of the pro-cesses involved in the vapour explosion, it will be checked if the temperature and pressure ranges of the applied tables are sufficient for realistic situations. Also the accuracy of the tables will be discussed and different tables will be compared.

Keywords: sodium cooled fast reactors, fuel-coolant interaction, thermo-dynamical properties

Mitja Uršič, Matjaž LeskovarJožef Stefan InstituteJamova cesta 39, SI-1000 Ljubljana, [email protected], [email protected]

Renaud MeignenInstitut de Radioprotection et Sureté NucléaireBP 17 – 92262 Fontenay-aux-Roses Cedex, [email protected]


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S3-137

NPP Krško Severe Accident Management Guidelines Upgrade

Nuclear Power Plant Krško (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry re-sponse to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG’s). SAMG’s are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products.NEK new SAMG’s revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess

risk of core damage situation during shutdown operation.

Keywords: core damage, SAMG, PCFV, PAR, SFP

Mario Mihalina, Srđan Špalj, Bruno Glaser, Robi Jalovec, Gordan JankovićNuklearna elektrana Krško Vrbina 12, 8270 Krško, Slovenija [email protected] , [email protected], [email protected], [email protected], [email protected]



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S3-139

Containment Modelling with the ASTEC Code

ASTEC is an integral computer code jointly developed by Institut de Radi-oprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für An-lagen- und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and second-ary circuits of the nuclear power plants (NPP), and in the containment.The ASTEC code was used to model and to simulate NPP behaviour dur-ing a postulated station blackout accident in the NPP Krško, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions

and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat struc-tures were used to simulate outer containment walls and internal steel and concrete structures.Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines.The accident analysis was focused on containment behaviour, however the complete integral NPP analy-sis was carried out in order to provide correct boundary conditions for the containment calculation. Dur-ing the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant containment parameters, such as compartments pressures and temperatures, is going to be discussed in the paper.

Keywords: ASTEC code, severe accident, PWR, containment

Siniša Šadek, Davor GrgićFaculty of Electrical Engineering and Computing, University of ZagrebUnska 3, 10000 Zagreb, [email protected], [email protected]


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Kaliopa ManchevaRisk Engineering Ltd.10 Vihren Str., 1618 Sofia, Bulgaria [email protected]

S3-146

Main Aspects and Results of Level 2 PSA for KNPP WWER-1000/B320

The PSA Level 2 for Kozloduy NPP (KNPP) is an update of an older study with wider scope of analysis. The older study represented the status of the units up to 2001. The current PSA Level 2 is based on the PSA Level 1 and represents the status of the units up to 2007 year concerning the sys-tems and procedures included in PSA level 1 and status up to 2011 for the systems and procedures (e.g. SAMG) related to containment and severe accident aspects. The study is performed after the PSA level 1 has been

finished and approved by the customer. Compare to the older analysis all modes of operation for analyzed in PSA level 1 event groups as well Spent Fuel Pool accidents are investigated. The analysis consists of both deterministic and probabilistic analysis. As part of deterministic analysis a contemporary containment strength analysis and accident progression deterministic analysis using last version of MELCOR are performed. The probabilistic analysis contains of two part: Interface PSA and CET are calculated using Riskspectrum program code. Two types of models for CET have been devel-oped: one for conditional probabilities calculations and a set of simplified CET’s for each PDS group – for integral model. The purpose of the first model is to be able to perform quick calculations and for sensitiv-ity analyses as well. The simplified CET’s are used for integral calculation of the model. Source Term analysis is mainly based on the MELCOR analyses results. All characteristics of the releases have been defined, i.e. location, mass, energy of radionuclide groups and activity of the released isotopes (most important are reported only).The main goals of the study are to analyze the status of the containment, systems designed to prevent containment failure and operator action required under the severe accident and to give quantitative as-sessment of the risk parameter LERF (Large Early Release Frequency). This report will present main aspects, results, finding and conclusions of the analysis. Based on the re-sults, the effectiveness of the current means and SAMG are assessed qualitatively.

Keywords: PSA level 2, WWER-1000, CET



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S3-150

Calculation of hydrogen concentration in containment during LOCA accident

The hydrogen sources in containment during and after design basis LOCA accident are usually calculated as part of SAR Chapter 6 analyses. There is requirement to limit hydrogen volumetric concentration to 4%. That is basis for sizing of hydrogen recombiners. The hydrogen concen-tration was calculated for NPP Krsko using simple GOTHIC containment model to reproduce SAR data. The influence of both, original electric and new passive hydrogen recombiners, was analyzed. Both recombiners were able to limit hydrogen concentration within the concentration required by plant design, but electric recombiner was more successful in hydrogen removal.

Keywords: containment, hydrogen recombiner, LOCA, GOTHIC

Davor Grgić, Tomislav Fancev, Siniša ŠadekUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected], [email protected], [email protected]


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S3-153

Optimization of OPDT Protection for Overcooling Accidents

Overcooling accidents are typically resulting in power increase due to negative moderator feedback. There are more protection set points re-sponsible for terminating power increase. OPDT protection set point is typically protection from exceeding fuel centre line temperature due to reactivity and power increase. It is important to actuate reactor trip signal early enough, but to be able to filter out events where actuation is not nec-essary. Different concepts of coolant temperature compensation as part of OPDT set point protection were studied for decrease of feed water tem-perature accident and for small main steam line breaks from full power for NPP Krsko. Computer code RELAP5 mod 3.3 was used in calculation. The influence of different assumptions in accident description as well as nuclear core characteristics were described.

Keywords: OPDT protection, decrease of feed water temperature, steam line break

Vesna Benčik, Davor Grgić, Nikola Čavlina, Siniša ŠadekUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected], [email protected], [email protected], [email protected]



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S3-157

Short overview of PSA quantification methods, pitfalls on the road from approximate to exact results

Over time the Probabilistic Safety Assessment (PSA) models have become an invaluable companion in the identification and understanding of key nuclear power plant (NPP) vulnerabilities. PSA is an effective tool for this purpose as it assists plant management to target resources where the larg-est benefit for plant safety can be obtained. PSA has quickly become an established technique to numerically quantify risk measures in nuclear power plants. As complexity of PSA models increases, the computation-al approaches become more or less feasible. The various computational approaches can be basically classified in two major groups: approximate and exact (BDD based) methods. In recent time modern commercially available PSA tools started to provide both methods for PSA model quan-tification. Besides availability of both methods in proven PSA tools the usage must still be taken carefully since there are many pitfalls which can drive to wrong conclusions and prevent efficient usage of PSA tool. For example, typical pitfalls involve the usage of higher precision approxima-tion methods and getting a less precise result, or mixing minimal cuts and prime implicants in the exact computation method. The exact methods are sensitive to selected computational paths in which case a simple hu-man assisted rearrangement may help and even switch from computa-

tionally non-feasible to feasible methods. Further improvements to exact method are possible and desir-able which opens space for a new research. In this paper we will show how these pitfalls may be detected and how carefully actions must be done especially when working with large PSA models.

Keywords: PSA, fault tree, event tree, BDD, risk analysis

Reni BanovUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected]

Zdenko ŠimićUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected]

Davor ŠtercPolytechnic of ZagrebVrbnik 8, Zagreb, [email protected]


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S3-163

Simulation and Validation of a Full Flow and Mini Flow Verification Test in NEK with Best Estimate HPIS Model in APROS

A model of High Pressure Injection System (HPIS) of the Krško NPP (NEK) was built using the APROS – Advanced PROcess Simulation en-vironment. APROS is a multifunctional software for modelling and dy-namic simulation of various physical processes of different types of nu-clear power plants, including PWRs. APROS allows us to build complete models of whole plants, with a complete thermal-hydraulic, electrical and regulation systems including reactor kinetics. Such models could be used for various purposes like as the for new NPP design verification, possi-ble optimization of NPP systems, for training purposes as an engineering simulation tool, for verification and validation of plant abnormal (AOPs) and emergency (EOPs) operating procedures and also as a model for full scope simulator. The data used to describe the properties of the HPIS sub-system modelled in APROS, were the data describing Krško NPP and its 23rd cycle. Basis for data collections, nodalization, structure and simplifi-cations was NEK RELAP5\MOD3.3 Engineering Handbook [1].The goal was to create a detailed simulation model a HPIS in APROS code and validate it for simulating the safety injection system response during

design basis accident. To reach the goal few plant surveillance tests were simulated. Upon completion of modelling/nodalisation and documenting of the model in the APROS, the model validation has been made in accordance with surveillance NEK procedure. The full and mini flow surveillance tests during re-fuelling shutdown were simulated by APROS. The HPIS simulation model injects water from the Refuel-ling water storage tank (RWST) directly in to the reactor vessel and the cold legs. The two Safety Injection (SI) pumps taking suction from RWST and pumps water through subsystems lines via associated check and control valves. Valves are modelled with corresponding pressure drop as a function of the flow veloc-ity, pipes with corresponding wall resistance and pumps with performance curve to achieving the right subsystem flow characteristics. The comparison with testing data obtained from Krško NPP surveillance tests and simulation results shows good agreements and flows are within the allowable limits.Such verified and validated HPIS model will be added to the full NEK APROS model for the next step of verification and validation full model itself.

Keywords: APROS, High pressure injection system, Nuclear power plant Krško, Validation, Verification, Safety injection, thermal-hydraulics

Dejan SlovencZEL-EN d.o.oHočevarjev trg 1, 8270 Krško, Slovenia [email protected]

Ivica BašićAPOSS d.o.oRepovec 23b, 49210 Zabok, Croatia [email protected]

Luka ŠtrubeljGEN energija d.o.oVrbina 17, 8270 Krško, [email protected]


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Session 4 – Radioactive Waste Management and Decommissioning

S4-112 O. Purtov, A. Masko, V. VasilchenkoMain Results of Updated Decommission Conception of NPPs Operating in Ukraine 58

S4-132 A. Knapp, I. Levanat, D. Šaponja-MilutinovićIs Croatia Going to Build a Radioactive Waste Repository? 59

S4-133 I. Levanat, V. LoknerCroatian Capacity for Management of the NPP Krsko Radioactive Waste 60

S4-134 L. AlujevićThe Safety Assessment Framework Tool (SAFRAN) – Description, Overview and Applicability 61

Session 4: Radioactive Waste Management and Decommissioning (RWMD)58

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S4-112

Main Results of Updated Decommission Conception of NPPs Operating in Ukraine

Results of long-term planning analysis based on consideration of 6 pos-sible scenarios for the nuclear energy development with 15 years and 20 years life time extensions of operation of nuclear power units beyond 30 year provided by original design are presented in the updated decommis-sion conception of NPP’s operating in Ukraine.These characteristics of the two main options for NPP decommissioning deferred or immediate dismantling, which is close to the level of accept-ability with relative superiority variant of deferred dismantling, are pre-sented.The best option for NPP unit decommissioning as comparative analysis results is the option with deferred dismantling with 30 years endurance time. It can be taken as a basis for optimal strategies for NPP unit decom-mission design development.

Cost estimations for the decommissioning of WWER-440 and WWER-1000 reactor type units are pre-sented in the updated conception. The updated cost assessment for required annual payments with uniform accumulation costs to the De-commission Fund corresponding deferred dismantling variant with 20 years life time extension operation time is 98,2 mln US$ per year. This value is 3.61 % of the electricity generated by NPP’s in Ukraine and supplied to the wholesale electricity market of Ukraine in 2012 base year.

Keywords: Updated Decommission Conception, long-term planning analysis, nuclear power units, WWER-440, WWER-1000, decommission cost estimation, life time extension

Oleg Purtov, Alexander Masko, Victor VasilchenkoState scientific – engineering center of control systems and emergency response (SSEC CSER), Pr. Geroev Stalingrada, 64/56, Kiev, 04213,[email protected], [email protected], [email protected]

Session 4: Radioactive Waste Management and Decommissioning (RWMD) 59


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S4-132

Is Croatia Going to Build a Radioactive Waste Repository?

Site selection process for low and intermediate level radioactive waste repository in Croatia was ended in 1999, nominating Trgovska gora as the potential macrolocation for the facility. Feasibility of the Trgovska gora disposal project was analyzed in a number of studies prepared by APO Ltd. from the mid-nineties up to 2003. An affirmative, though preliminary and largely generic safety assessment was completed. Spe-cific microlocations were selected and analyzed based on literature data (garnished with low-resolution digital satellite pictures), and the best microlocation was tentatively narrowed down to Pavlovo brdo.After 2003, no further activities related to the repository project were undertaken for nearly ten years, until in its public procurement plan

for 2013 the Croatian Fund for financing the NPP Krško decommissioning and waste disposal dedicated over half a million euro to continuation of the project.In general, safe radioactive waste disposal pre-requires establishment of a complex national framework with appropriate functionality and competence; with such a framework established, decisive first steps towards building a repository are to identify potentially suitable locations and to ensure local community consent and cooperation. The rest should mainly be routine. But in Croatia, both lack of proper frame-work and the project history of indecisiveness may adversely affect further developments.Trgovska gora was designated as the potential location in the national land use plan only after three other potential locations had been dismissed by political decisions based on the largely assumed adverse attitudes of local communities. Repository project now appears to depend on cooperation of a single lo-cal community hosting the only potential site. The site has never been visited by any repository project participants, nor has the local community ever been officially contacted in an open and straightforward way, despite the 20-year old history of the project activities.Of course, the local community is not entirely unaware of the wavering interest of the state in building the repository there.Under such circumstances, considerable effort, skill and knowledge will be needed to establish confi-dence of that community in the authorities and their intentions and competence – and yet all that may not be sufficient.

Keywords: radioactive waste repository, Croatia, Trgovska gora

Alemka Knapp, Ivica Levanat, Diana Šaponja-MilutinovićPolytechnic of ZagrebVrbik 8, 10 000 Zagreb, [email protected], [email protected], [email protected]

Session 4: Radioactive Waste Management and Decommissioning (RWMD)60

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S4-133

Croatian Capacity for Management of the NPP Krsko Radioactive Waste

Together with Slovenia, Croatia is responsible for management of the NPP Krško radioactive waste and spent fuel. So far, no firm agreement on specific solutions has been reached between the two countries. On the contrary, all activities related to revision and development of the joint Program of the NPP decommissioning and spent fuel and radioac-tive waste management were discontinued several years ago.Unless Slovenia and Croatia definitely agree on joint solutions in the meantime, Croatia will have to begin transfer of one half of the NPP Krško spent fuel and radioactive waste into its territory in about nine years. Presently, however, Croatian capacities for such an undertaking are seriously inadequate in several respects, and they are not developing in any promising way.

For no rational reason, this state of the Croatian capacities has been maintained (recently even deterio-rating in some respects) for at least a decade, disregarding also for at least two years the explicit require-ments of the EU Directive 2011/70/EURATOM aimed at establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste. In fall of 2012, Fund for financing the NPP Krško decommissioning and spent fuel and radioactive waste disposal was appointed as the Croatian expert organization for revision and development of the program of these same activities it had originally been supposed only to finance. In 2013 the Fund expanded its activities to a low-profile attempt at revitalization of the Croatian radioactive waste repository project, although the Fund is not yet properly capacitated for either of these tasks. The above is hardly in compliance with the Directive requirements, such as to establish “a national legisla-tive, regulatory and organisational framework” “that allocates responsibility... between relevant compe-tent bodies”. The lack of competent experts in the field appears to affect the quality of Croatian legislative and regulatory capacity as well. That is clearly seen in the recently amended Croatian basic law on nuclear safety and radiation protection, which actually presents a step backward in the radioactive waste disposal regulation, although this should have been its prime concern in the present circumstances.

Keywords: radioactive waste, Croatian national framework, NPP Krsko, EU Directive 2011/70

Ivica Levanat Zagreb PolytechnicVrbik 8, 10 000 Zagreb, Croatia [email protected], [email protected]

Vladimir Lokner,APO Ltd.Savska 41/IV, 10 000 Zagreb, Croatia [email protected]

Session 4: Radioactive Waste Management and Decommissioning (RWMD) 61


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S4-134

The Safety Assessment Framework Tool (SAFRAN) – Description, Overview and Applicability

The SAFRAN tool (Safety Assessment Framework) is a user-friendly software application that incorporates the methodologies developed in the SADRWMS (Safety Assessment Driven Radioactive Waste Manage-ment Solutions) project.The International Atomic Energy Agency (IAEA) organized the Inter-national Project on Safety Assessment Driving Radioactive Waste Man-agement Solutions (SADRWMS) to examine international approaches

to safety assessment for predisposal management of all types of radioactive waste, including disused sources, small volumes, legacy and decommissioning waste, operational waste, and large volume natu-rally occurring radioactive material residues.SAFRAN provides aid in:

• Describing the predisposal RW management activities in a systematic way, • Conducting the SA (safety assessment) with clear documentation of the methodology, assump-

tions, input data and models,• Establishing a traceable and transparent record of the safety basis for decisions on the proposed

RW management solutions,• Demonstrating clear consideration of and compliance with national and international safety stand-

ards and recommendations.The SAFRAN tool allows the user to visibly, systematically and logically address predisposal radioactive waste management and decommissioning challenges in a structured way.It also records the decisions taken in such a way that it constitutes a justifiable safety assessment of the proposed management solutions.The objective of this paper is to describe the SAFRAN architecture and features, properly define the terms safety case and safety assessment, and to predict the future development of the SAFRAN tool and assess its applicability to the construction of a future LILW (Low and Intermediate Level Waste) storage facility and repository in Croatia, taking into account all the capabilities and modelling features of the SAFRAN tool.

Keywords: SAFRAN, SADRWMS, safety case, safety assessment, radioactive waste management, predis-posal, decommissioning

Luka AlujevićHEP d.d.Ulica grada Vukovara 37, Zagreb, [email protected],

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62

Session 5 – Operation and Maintenance Experience

S5-102 T. Tomašić, I. Vuković, A. BakićARCHER – Advanced System for RPVH Inspection and Repair 64

S5-103 M. Vavrous, K. Markulin, M. BrekaloIn-service inspection of primary circuit components: Qualification of non-destructive testing applications 65

S5-104 M. Kekelj, M. Budimir, N. Pavlović, R. GracinMultielement Ultrasound and Eddy Current Integrated Probe for Non-destructive Evaluation of Nuclear Reactor Pressure Vessel Head Penetrations 66

S5-105 P. MateljakCASTOR – Advanced System for VVER Steam Generator Inspection 67

S5-107 D. Borović, I. VukovićGIMIS – Integral Solution for the In-Service Inspection Management of Components in NPPs 68

S5-118 A. Bakić, M. Pajnić, T. TomašićNew Solution For Ultrasonic Pipe Inspection System 69

S5-124 Z. Šimić, R. BanovDevelopment of the Operational Events Groups Ranking Tool 70

S5-129 S. ReškovićAutomatic Analysis for VVER and PWR 71

S5-136 B. Bach, R. Čižmek, B. BožinDeveloping Effective Corrective Action Plan in Krško NPP 72


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S5-138 M. Chambers, B. Kurinčić16x16 Vantage+ Fuel Assembly Flow Vibrational Testing 73

S5-140 D. DjakovićControl of Special Processes at Krško NPP 74

S5-142 I. Vrbanić, I. BašićOn Use of PSA for Characterization of Risk Significance of Operational Events and Issues in NPPs 75

S5-161 M. Gluhak, M. SenegovićShutdown Safety in NEK 76

S5-206 M. DudašOperator Fundamentals 77

Session 5: Operation and Maintenance Experience (OME)64

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S5-102

ARCHER – Advanced System for RPVH Inspection and Repair

The reactor pressure vessel head (RPVH) is an integral part of the reac-tor coolant pressure boundary. Its integrity is important for the safe and reliable operation of the nuclear power plant (NPP). After detection of the leakage and cracks in French NPP, followed by another that occurred in NPP in USA, methods and frequency of inspection were defined, and are strictly regulated by the US NRC Order EA-03-009 (substituted lately by ASME Code Case N-791-1) since 2003.Usual scope of inspection from inner side of RPVH comprises of visual inspection of the surface, ultrasonic testing (UT) and eddy current test-ing (ET) of the penetration nozzle and ET of the J-groove weld and nozzle outside surface below the weld.ARCHER, new INETEC’s manipulator, is designed to provide full scope inspection of the RPVH, by use of various test modules and by perform-

ing the surface repair action on J-groove weld. It is adjustable to work with different types of penetration nozzles and thermal sleeves on both VVER and PWR type of NPP. Due to complex geometry each module is specially designed for particular type of examination. Modules are exchanged through the docking sys-tem without need for personnel to enter under the head region, thus reducing the personnel’s exposure to the ionizing radiation.The end effectors are used to determine the surface flaws or cracks on inner diameter surface of penetra-tion nozzle gap. It guides a slim sword-like probe which carries a pair of TOFD transducers for detection and sizing of circumferential and axial cracks, an eddy current cross-wounded coil, and a zero-degree UT probe through a gap between the penetration nozzle and thermal sleeve. In case of a non-sleeved penetra-tion nozzle, an open housing module is used.J-groove module is designed to fit geometry of the J-groove weld of penetration nozzle, vent pipe and funnel guide. The whole weld area (2” mm on shell side and ½” on nozzle side) is covered by two specially designed array eddy current probes.Surface flaws, discovered by eddy current examination of J-groove weld, define the scope of the automat-ed surface repair module (ASR module) performed by the grinding method. Specially developed grinding procedure, based on the surface probing and UT results, ensures the treatment doesn’t affect the origi-nally designed structural integrity basis. When compared with the other repairing methods, ASR module significantly reduces the inspection time and radiation exposure of the personnel, and does not introduce residual stress into the structural material.The paper describes the system’s capabilities and features, and its advantages compared to other systems for performing the RPVH inspection and repair activities on PWR and VVER reactors.

Keywords: RPVH, primary cycle

Tomislav Tomašić, Igor Vuković, Ante BakićINETEC – Institute for Nuclear TechnologyDolenica 28, 10253, Lučko, [email protected], [email protected], [email protected]

Session 5: Operation and Maintenance Experience (OME) 65


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S5-103

In-service inspection of primary circuit components: Qualification of non-destructive testing applications

Reliability of operation of nuclear power plant heavily depends on regular maintenance that is performed during annual outages. In order to preserve environmental safety and integrity of reactor coolant circuit, primary cir-cuit components are specially addressed in maintenance programs and special requirements are applied for these components. Non-destructive testing in-service inspections are performed on primary circuit compo-nents to determine and assess its condition without affecting the tested equipment.Qualification of in-service inspection is requirement introduced to ensure that applied in-service activity has capabilities to meet prescribed require-ments. This is proved through a number of tests that provide both theo-

retical and practical evidence about capabilities of selected non-destructive testing methods.Important aspect of qualification is component for which inspection is being qualified and applied stand-ards. Depending on the applied standards and codes (ASME, ENIQ, GOST, etc.) and it’s requirements, qualification approach varies significantly and results with different type and number of tests. While cer-tain approaches place more emphasize on practical trials other approaches require more documentation and statistical evidence. This paper will present latest efforts and overview of achieved results of INETEC in field of qualification of in-service inspections of Reactor Vessel Closure Head (RVCH) of Pressurized Water Reactor (PWR) types, VVER-440 Steam generator tubing, VVER-1000 Steam Generator tubing and VVER-440 steam generator collector welds.

Keywords: non-destructive testing, qualification, in-service inspection, reliability, primary circuit compo-nents

Matija Vavrous, Krunoslav Markulin, Marijan BrekaloINETEC-Institute for Nuclear technology Dolenica 28, 10250 Lucko, [email protected], [email protected]


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S5-104

Multielement Ultrasound and Eddy Current Integrated Probe for Non-de-structive Evaluation of Nuclear Reactor Pressure Vessel Head Penetrations

An integrated combined UT (ultrasound testing) and ECT (eddy current testing) multielement probe prototype for inspection of gaps in nuclear reactor pressure vessel head penetrations (RPVHPs) was numerically modelled, designed, assembled and tested in this work. The probe pro-totype consists of an acoustically active head and an acoustically passive body. The active part consists of an axial and a circumferential UT TOFD (time of flight diffraction) configuration pairs of central frequency of 6.2 MHz, of a normal beam 0° single element disc shape UT probe of central frequency of 2.25 MHz and a cross wound ECT probe. All these elements have been integrated in the polymer probe head of dimensions 29mm x 24.4 mm x 2.8 mm, with the OD (outer diameter) of 69.85 mm. The head is mounted on a flexible stainless steel probe body, able to follow the com-plicated structure of narrow gaps between penetration pipes and thermal

sleeves. The prototype was assembled according to numerical optimizations results, and its electrome-chanical properties were tested on a stainless steel calibration block. The modelled and experimental results of electromechanical probe signals showed excellent agreement and all the artificially made indi-cations on the calibration block were successfully detected and sized.

Keywords: NDT, RPVHP, UT, ECT

Matija Kekelj, Marko Budimir, Nikola Pavlović, Renato GracinINETEC – Institute for nuclear technology Ltd. Dolenica 28, Lučko, [email protected], [email protected], [email protected], [email protected]



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S5-105

CASTOR – Advanced System for VVER Steam Generator Inspection

From the safety point of view, steam generator is a very important compo-nent of a nuclear power plant. Only a thin tube wall prevents leakage of ra-dioactive material from the primary side into the environment. Therefore, it is very important to perform inspections in order to detect pipe damage and apply appropriate corrective actions during outage. Application of the nondestructive examination (NDE) technique, that can locate degrada-tion and measure its size and orientation, is an integral part of nuclear power plant maintenance.

The steam generator inspection system is consisted of remotely controlled manipulator, testing instru-ment and software for data acquisition and analysis. Recently, the inspection systems have evolved to a much higher level of automation, efficiency and reliability resulting in a lower cost and shorter outage time. Electronic components have become smaller and deal with more complex algorithms. These sys-tems are very fast, precise, reliable and easy to handle. The whole inspection, from the planning, exami-nation, data analysis and final report, is now a highly automated process, which makes inspection much easier and more reliable.This paper presents the new generation of INETEC’s VVER steam generator inspection system as ulti-mate solution for steam generator inspection and repair.

Keywords: NDE, UT, ET, VVER, steam generator

Petar MateljakINETEC – Institut za nuklearnu tehnologiju d.o.o.Dolenica 28, 10250 Zagreb, [email protected]


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S5-107

GIMIS – Integral Solution for the In-Service Inspection Management of Components in NPPs

Performance of in-service testing and inspection of components and sys-tems in nuclear power plants are required in order to maintain the nuclear power plant while in operation and to return the plant to service, follow-ing plant outages.GIMIS is comprehensive software that integrates all processes, functions and data related to planning, administrating and executing inspections on systems, structures and components in nuclear power plants. The soft-ware is designed as a web application developed using the Microsoft ASP.NET technology, database is a Microsoft SQL, and client is composed of JavaScript frameworks. It can be adapted to local language, regulations,

and requirements according to the power plant needs.The application consists of seven interdependent modules, namely: components, equipment, personnel, requirements, inspection planning, inspection execution, and reports.GIMIS deals with component, equipment and personnel management (both internally employed and outsourced), also provides full component history including uploaded documentation, drawings, pre-vious inspection results, and supports various types of requirements. It enables scheduling facilitation using component data and compliance requirements, offers generation of all inspection relevant docu-mentation and reports, and covers the equipment management including calibration requirements, cer-tification of equipment and allocation to specific inspections/outages, as well as personnel certifications and allocation to specific inspections/outages.The paper describes the content and functionality of the GIMIS application and provides information of its built-in capabilities and features.

Keywords: in-service inspection, nuclear power plant, inspection management, information system, soft-ware

Dominik Borović, Igor VukovićINETEC – Institute for Nuclear TechnologyDolenica 28, HR-10250 Zagreb-Lučko, [email protected], [email protected]



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S5-118

New Solution For Ultrasonic Pipe Inspection System

Ultrasonic pipe inspection system comparing to visual inspection system requires much more drive power due to large cable bundle and resistance that occurs as friction between the probes and inner pipe surface. In ad-dition, the instrument is usually outside of the pipe especially for small diameter pipes which means lot of signal wires together with motor power supply. This is very adverse for the good quality of the measuring signal causing a limitation on the cable length. This paper describes in detail the new ultrasonic pipe inspection system with a unique driving mechanism. The driving mechanism consists of three pneumatic grippers connected together with a flexible lead screw.

The first gripper carries a high resolution camera, the middle one has a central screw nut while the driv-ing motor is mounted on the last one. The movement sequence begins when the first and last grippers are locked, then turning the lead screw, gripper with a nut is moving forward. After that the middle gripper is locked and others two are released, rotating the lead screw in opposite direction all system including probes and a cable now is moved forward. Repeating the sequence of these two steps the system moves through the pipe with the speed up to 50 mm/s. The system is modular and can be easily adapted for dif-ferent pipe diameters between 80 and 600 mm.Driving mechanism is respectively followed by a module for incremental rotation, module with probes and waterproof control unit. It is very important that electronic amplifiers are located very close to the motors. In this way, the current fluctuations in the rest of the cable is minimized which is beneficial to re-duce the noise in the measuring signal. Each module is equipped with guiding wheels as well as full length of the cable reducing translation resistance and smooth passing through the bends and over any changes in the cross section of the pipe. In case of system failure it is designed to stay in the position which is suit-able for manually pulls out from the tube. PC software makes the system highly automated and easy to operate. Due to large flexibility and high driving torque large sections of pipe of several hundred meters in length with many bends and vertical sections can be inspected.

Keywords: ultrasonic pipe inspection system, flexible screw drive

Ante Bakić, Mladen Pajnić, Tomislav TomašićINETEC Institute for Nuclear TechnologyDolenica 28, 10000 Zagreb, [email protected]


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S5-124

Development of the Operational Events Groups Ranking Tool

Both because of complexity and ageing, facilities like nuclear power plants require feedback from the operating experience in order to further im-prove safety and operation performance. That is the reason why signifi-cant effort is dedicated to operating experience feedback.This paper contains description of the specification and development of the application for the operating events ranking software tool. Robust and consistent way of selecting most important events for detail investiga-tion is important because it is not feasible or even useful to investigate all of them. Development of the tool is based on the comprehensive events characterisation and methodical prioritization. This includes rich set of events parameters which allow their top level preliminary analysis, differ-ent ways of groupings and even to evaluate uncertainty propagation to the ranking results. One distinct feature of the implemented method is that user (i.e., expert) could determine how important is particular ranking parameter based on their pairwise comparison.For tools demonstration and usability it is crucial that sample database is

also created. For useful analysis the whole set of events for 5 years is selected and characterised. Based on the preliminary results this tool seems valuable for new preliminary prospective on data as whole, and especially for the identification of events groups which should have priority in the more detailed assess-ment. The results are consisting of different informative views on the events groups importance and related sensitivity and uncertainty results. This presents valuable tool for improving overall picture about specific operating experience and also for helping identify the most important events groups for further assess-ment. It is clear that completeness and consistency of the input data characterisation is very important to get full and valuable importance ranking.Method and tool development described in this paper is part of continuous effort of the European Clear-inghouse of Operating Experience Feedback for Nuclear Power plants operated by the European Com-mission – Joint Research Centre – Institute for Energy and Transport (EC JRC-IET).

Keywords: operational experience feedback, safety, tool, events characterisation, event groups ranking

Zdenko Šimić, Reni BanovEC JRC-IET-NRSA, European Commission Joint Research Centre-Institute for Energy and Transport-Nuclear Reactor Safety AssessmentPostbus 2, 1755 ZG, Petten, The [email protected]

University of Zagreb Faculty of Electrical Engineering and ComputingUnska 3, 10000 [email protected]



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S5-129

Automatic Analysis for VVER and PWR

EddyOne Analysis software developed by INETEC includes module for automatic detection of defects in eddy current data for VVER and PWR type of plants. Automatic analysis comprises various stages in order to successfully and efficiently detects defects in the signal, such as pre-pro-cessing of the signal data, initial detection, classification, characterization and reporting. Detecting defects in eddy current signal is complicated

task because typical signal often contains lots of data points representing regular signal and noise, with only fraction of the total signal being actual defects. This poses potential problems for the automatic analysis algorithms which should be able to find all the defects in the signal and at the same time have as little detection of false indications as possible while still being relatively fast. In order to achieve this, it can be helpful to have some sort of initial detection algorithm which can relatively quickly discard parts of the signal that does not contain defects under some probability. Such algorithm would then set ground for more powerful and time consuming algorithms to process the remaining of the signal which might have potential defects. There are many types of the defects in the steam generator tubes characterized by the different degradation mechanisms. These defects can sometimes significantly differ from one to another so we adopted the detection algorithm for defect individually. These specialized algorithms yield better results in terms of time and accuracy. Neyman-Pearson (N-P) detector which maximizes probability of detecting actual defects in the presence of noise for a given probability of false detection is used as the basis for the initial detection. By using N-P detection, we have achieve reduction of up to 60% in the total data set, meaning that subsequent algorithms will have to analyse about 40% of the remaining data. N-P detector was further improved by adjusting it for different types of defects. After initial detection classifi-cation of defects was performed by implementing rules based classification system for each type of defect. The last step is measuring defect and reporting results.

Keywords: eddy current, automatic analysis, steam generator

SAŠA REŠKOVIĆINETEC – Institute for Nuclear TechnologyDolenica 28, Zagreb, [email protected]


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S5-136

Developing Effective Corrective Action Plan in Krško NPP

Experience shows that many events could have been prevented if lessons had been learned from previous incidents. Event reporting thus has be-come an increasingly important aspect of the operation and regulation of all safety-related and public health industries. Different industries such as aeronautics, chemicals, transport and of course nuclear depend on Op-erating Experience (OE) feedback programs to provide lessons learned about safety. The information available under an OE programme for these organizations comprises internal event reports and analysis and external operating experience including reports on low level events and near miss-es and other relevant operating performance information.

The worldwide OE programme (such as WANO OE) in nuclear power plants provides opportunity to learn from events at other plants. In particular, it alerts plants to mistakes or events that have occurred at other nuclear power plants and enables them to take corrective actions to prevent similar occurrences at their own plant. The intent of the effective and efficient OE program is therefore to improve personnel/plant safety, reliability and commercial performance of the operating nuclear power plants. Such a pro-gramme ensures that operating experience is analysed, events important to safety are reviewed in depth, lessons learned are disseminated to the staff and to the relevant national and international organizations and corrective actions are effectively implemented. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures.Krško NPP is developed its own OE program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The OE is a part of the Corrective Action Pro-gram (CAP), which is among top management programs.The purpose of this article is to present a part of the Krško NPP OE program regarding to developing ef-fective and efficient corrective action plan based both on internal and industry events. It will describe the key steps of developing, implementing and assessing the effectiveness of corrective actions that address issues identified in the OE program and includes the following aspects:

• a. Developing effective corrective actions,• b. Prioritizing corrective actions• c. Implementing corrective actions successfully• d. Assessing the effectiveness of the corrective actions.

Keywords: Event, Operating Experience, Corrective Action

Bruno Bach, Rudi Čižmek, Bojan BožinNuclear Power Plant KrškoVrbina 12, Krško, [email protected], [email protected], [email protected]



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S5-138

16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

Nuklearna Elektrarna Krško (NEK) has experienced leaking fuel after in-creasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting.NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mix-

ing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down…), operational cycle duration increase from 12 to 18 months (increas-ing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate).The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier.The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature and pressures. The aim of vibrational testing was to determine a vibration mechanism that would explain the fretting behav-iour such as grid vibration, fuel rod vibration or FA vibration, however this has proven elusive.

Keywords: Vibration, Fuel Assembly, Fuel Reliability, Fretting

Martin Chambers, Bojan KurinčićNuklearna Elektrarna KrškoVrbina 12, 8270 Krško, [email protected], [email protected]


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S5-140

Control of Special Processes at Krško NPP

Control of Special Processes is one of the 18 criteria found in 10CFR50 Appendix B – Quality Assurance Criteria for Nuclear Power Plants. Weld-ing, brazing, non-destructive testing, coating and heat-treating are the main special processes that have to be controlled and results of which are highly dependent on the control of the process or the skill of opera-tors and in which the specified quality cannot be readily determined by

inspection or test of the process item. Almost all plants are in the process of extending plant operating licenses and special processes play a critical role in establishing conditions and providing reliable data for the technical justification to extend operations.At Krško NPP the Quality Control Department (QC) provides Control of Special Processes as part of Quality and Nuclear Oversight Division (SKV). In light of controlling special processes, the QC Depart-ment focuses on the control of welding and heat-treating processes in quality-affecting activities, using non-destructive examinations (NDE) being implemented and applicable to our personnel.Special processes at Krško NPP are accomplished utilizing qualified personnel, qualified procedures and qualified equipment, and comply with the requirements of applicable codes and standards. Personnel, equipment and procedures, used for welding in ASME Code applications, are qualified in accordance with requirements of ASME Section IX. Specifications and procedures for these activities shall be ap-proved by the welding engineers and quality organizations having jurisdiction. NDE personnel are quali-fied and certified in accordance with ANSI/ASNT CP-189 (Standard for Qualification and Certification of NDE Personnel), considering qualifications in accordance with ASME Sections III, V, IX and XI, and ASME B31.1 Power Piping. This paper describes Control of Special Processes at Krško NPP in light of the aspects mentioned above, non-destructive examinations, mostly used for control of welding processes according to corrective and preventive program requests, as well as its importance in the realization of modification projects in the form of an inspection survey of the performance of dedicated and approved inspection plans. Criterion 10 of 10CFR50 Appendix B – Inspection – is closely associated to all QC activities and together with Cri-terion 9 – Control of Special Processes – defines quality control scope and organization. A good example of the application of control of special processes and participation of the QC Department in modification projects was the manufacturing and installation of PCFVS (Passive Containment Filtered Venting Sys-tem) as a part of Safety Upgrade Program at Krško NPP.

Keywords: special processes, quality control, nuclear oversight, welding, non-destructive examinations

Dragoslav DjakovićNPP KrškoVrbina 12, Krško 8270, [email protected]



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S5-142

On Use of PSA for Characterization of Risk Significance of Operational Events and Issues in NPPs

Like in the case of other industrial facilities, different low-level operation-al events or issues can be observed to take place during the operation of nuclear power plants. Similarly, a number of findings and observations is being made during the safety reviews which are regularly conducted at the utilities or at the regulatory side. In most of those cases, there is no any consequence concerning the nuclear safety or risk. The philosophy established during the previous two decades regarding the operating ex-perience feedback is that any of those observed occurrences or issues may be an opportunity to learn and further improve plant safety performance.

In order to distinguish or “filter out” those events or issues which deserve a deeper consideration, a con-cept of “risk significance” is being introduced into the field of operational events and issues investigation. Paper presents and discusses the probabilistic safety analysis (PSA) concepts for the characterization of operational events and safety issues in terms of quantitative risk significance. The types of information from a plant-specific PSA which can be used for this purpose are pointed out and examples and illustra-tions provided. Basic approaches which can, also, be used by those evaluators or investigators which are not necessarily specialized in PSA are outlined.

Keywords: PSA, nuclear power plants, operational events, risk significance

Ivan Vrbanić, Ivica BašićAPoSS d.o.o.Repovec 23b, HR-49210 Zabok, [email protected], [email protected]


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S5-161

Shutdown Safety in NEK

Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage.Back in 1990s, the industry adopted guidance to meet safety require-ments by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those re-

quirements. However, in the last decade inadequate “defence in depth” has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: “The special char-acteristics of nuclear technology are taken into account in all decisions and actions. Reactivity control, continuity of core cooling, and integrity of fission product barriers are valued as essential, distinguishing attributes of nuclear station work environment.” NEK has recognized the importance of “defence in depth” concept and has implemented various ways to keep safety margins adequate. Strive to stand among the best in the industry, in terms of safety, calls for constant improvement, innovation and implementation of new concepts. The areas such as outage scheduling, risk assessment, procedures, equipment protection and error prevention behaviour are the main areas NEK is building its defence upon.

Keywords: outage, shutdown, safety, protection

Mario Gluhak, Marko SenegovićKrško Nuclear Power PlantVrbina 12, 8270 Krško, [email protected], [email protected]



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S5-206

Operator Fundamentals

Recent events show the need for constant attention on the operator fun-damentals, in the commercial nuclear industry. The first report about de-cline in the application of operator fundamentals during plant operational activities and transient situations was issued in July 2005. Analyses of the events recorded during 18 month period between 2010 and 2011 show similar causes and contributors like it was before July 2005. Due to that fact, the WANO issued SOER 2013-1 Operator Fundamentals Weakness-

es with proposed suggestions how to analyse area of operator fundamentals and gives recommendations for effective and sustainable corrective actions.Operator fundamentals are the essential knowledge, skills, behaviours, and practices that operating crews need to apply to operate the plant effectively.These fundamentals are as follows:

• Monitoring plant indications and conditions closely• Controlling plant evolutions precisely• Operating the plant with a conservative bias• Working effectively as a team• Having a solid understanding of plant design, engineering principles, and sciences

NEK analysed area of operator fundamentals and verified how consistently the basic principles in the plant control are followed in practice. Some opportunities for improvement were recognized for the training area, operational procedures format improvement and improvement in process of preparation of the planned activities during power operation or during plant shutdown. Among other measures, stabil-ity in operation with a sufficient safety margin can be achieved only through continuous monitoring of the operational practice and by constant highlighting of the operational standards.

Keywords: Operator fundamentals, Knowledge, Conservatism, Control, Team work

Mladen DudašKrško Nuclear Power PlantVrbina 12, 8270 Krško, Slovenia [email protected]

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78

Session 6 Environment, Public Relations and Safety Culture

S6-113 B. Manchev, V. Yordanova, B. NenkovaConfiguration Management Program – a part of Integrated Management System 79

S6-115 K. Trontl, D. Pevec, M. Matijević, R. Ječmenica, J. LebegnerPublic Opinion Survey – Energy – The Present and the Future – 2012/2013 81

S6-123 R. Bišćan, I. FifnjaKrško NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System) 82

S6-131 I. Jakić, R. FilipinAnalysis of Public Opinion Survey “Nuclear Energy – the Present and the Future” (2000–2012) 84

S6-147 S. Pleslić, G. Jimenez VarasKnowledge Loss Risk Assessment in Education and Industry 85

S6-158 I. Prlić, M. Surić Mihić, P. Shaw, M. Hajdinjak, Ž. Božina, D. Kosmina, Z. Cerovac

EU Outer Borders and Radiation – An urgent need for standardization, new detector technologies and education harmonization 86

S6-159 I. Prlić, M. HajdinjakNORM – Radionuclide transfer studies- A modern approach according Directive 2013/59/EURATOM 87

S6-160 Z. ŠimićOn the Impact to the Human Health from the Fukushima Nuclear Accident 88

Session 6 - Environment, Public Relations and Safety Culture (EPRSC) 79


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S6-113

Configuration Management Program – a part of Integrated Management System

The recently issued International Atomic Energy Agency (IAEA) publica-tions (GS-R-3, GS-G-3.1 and GS-G-3.5) regarding Management Systems for Facilities and Activities define requirements for creation, introduc-tion, evaluation and continuously improvement of the Management Sys-tem, which unifies the safety, health, environment, security, quality and economic elements. According to GS-R-3 the Integrated Management System is based on defined processes identified in the enterprises: Man-aging, Basic and Supporting processes.At implementation of their activities, the organizations often apply other standards in their interrelations with suppliers and the parties concerned – ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007, regarding quality, environment and occupational health and safety management. The integration of the standards of both series ensure the observance of the common management principles that reflect the best practices of

management as leadership, participation of the people, process approach, continuously improvement, systematical approach to the management and approach based on facts used at the making decisions. The main objective of the Integrated Management System introduction is to ensure safety considering the influence of all additional impacts taken together.The Integrated Management System is based on the process approach at implementation of the activities in nuclear power plant. The transition to the process oriented approach require long period of time, dur-ing which the distribution of the responsibilities is optimized up to the level that will satisfy the require-ments, reach and maintain the stipulated objectives.The Configuration Management (CM) is an integrated management process by means of which conform-ity between design requirements, physical configuration and the plant documentation is ascertained and maintained during the entire life cycle of the facility. Processes within configuration management are not isolated, but are part of the Integrated Management System.CM ensures that during the entire operational life of the plant the following requirements are met:

• The basic design requirements of the plant are established, documented and maintained;• The physical structures, systems and components (SSCs) of the plant are in conformity with the

design requirements;• The physical and functional characteristics of the plant are correctly incorporated in the opera-

tional and maintenance documentation, as well as in the documents for testing and training;

Bogomil Mancev, Vanja YordanovaRisk Engineering Ltd10 Vihren Str., 1618 Sofia, [email protected], [email protected]

Boyka NenkovaGSR Ltd10 Vihren Str., 1618 Sofia, [email protected]

Session 6 - Environment, Public Relations and Safety Culture (EPRSC)80

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• The changes in the design documentation are incorporated in the physical configuration and• the operative documentation;• The changes in the design are minimized by management process for review according to approved

criteria.The purpose of this report is to try to clarify the place of configuration management program within the Integrated Management System of Kozloduy NPP and to present the computerized information system for organization of the operational activities (IS OOA) as a tool for effective management of the facility.

Keywords: integrated management system, configuration management



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S6-115

Public Opinion Survey – Energy – The Present and the Future – 2012/2013

During the academic year 2012/13 the Department of Applied Physics of the Faculty of Electrical Engineering and Computing, University of Za-greb conducted a public opinion survey entitled “Energy – The Present and the Future” among student population of more than 950 individuals. The tested population consisted of the University of Zagreb five faculties’ and one polytechnic school’s students: the Faculty of Electrical Engineer-ing and Computing, the Faculty of Food Technology and Biotechnology, the Faculty of Chemical Engineering and Technology, the Faculty of Civil Engineering, the Faculty of Science, and the Polytechnic of Zagreb. The questions in the survey covered several different energy issues, including the present and the future energy resources, the acceptability of different fuel type power plants, the environmental protection and global warm-ing, the radioactivity, the waste issues, reliable information sources, and position of participants towards climate change issues, , as well as Euro-pean Union and Croatian goals set for the year 2020. The basic results of survey analysis for nuclear oriented questions, as well as the comparison of results of the current survey with the results of the similar surveys con-ducted in the academic years 2007/08 and 2009/2010, are reported in this paper.

Participants generally express high level of formal environmental awareness. However, their choices and attitudes are in a contradiction to claimed eco-orientation, as well as to the scientific facts. The discrep-ancies are particularly noticeable in parts of the survey dealing with the nuclear energy and the nuclear power plants. The participants are also demonstrating lack of knowledge on nuclear issues especially regarding radioactive waste management, as well as economics and operational safety of nuclear power plants.

Keywords: public opinion survey, energy, nuclear energy

Krešimir Trontl, Dubravko Pevec, Mario Matijević, Radomir JečmenicaUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected], [email protected], [email protected], [email protected]

Josip LebegnerHEP d.d.Ulica grada Vukovara 37, 10000 Zagreb, [email protected]


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S6-123

Krško NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System)

Nuklearna Elektrarna Krško (NEK) has undertaken Nuclear Safety Up-grade Projects as a safety improvement driven by the lessons learned from the Fukushima-Daiichi Accident. Among other projects, new modifica-tion 1008-VA-L Passive Containment Filtered Vent (PCFV) System has been installed which acts as the last barrier minimizing the release of radioactive material into the environment in case of failure of all safety systems, and to insure containment integrity during beyond design basis accidents (BDBA). In addition, modification 1002-GH-L Severe Accident

Hydrogen Control System (PAR) has been implemented to prevent and mitigate the consequences of explosive gas generation (hydrogen and carbon monoxide) in case of reactor core melting. To ensure containment integrity for all design basis accidents (DBA) and BDBA conditions, NEK has eliminated existing safety-related electrical recombiners, replaced them with two safety-related passive autocatalytic recombiners (PARs) and added 20 new PARs designed for the BDBA conditions.Krško NPP Quality Assurance Plan has been applied to Nuclear Safety Upgrade Projects (PCFV System and PAR System) through the following activities:

• Internal audit of modification process was performed.• Supplier audits were performed to evaluate QA program efficiency of the main design organization

and engineering organizations.• Evaluation and approval of Suppliers were performed.• QA engineer was involved in the review and approval of 1008-VA-L and 1002-GH-L modification

documentation (Conceptual Design Package, Design Modification Package, Installation Package, Field Design Change Request, Problem/Deficiency Report, and Final Documentation Package).

• Purchasing documentation for modifications 1008-VA-L and 1002-GH-L (technical specifications, purchase orders) has been verified and approved by QA.

• QA and QC engineers were involved in oversight of production and testing of the new 1008-VA-L and 1002-GH-L plant components.

• Implementation of modifications was verified and approved through continuous QA and QC in-volvement (modification team members).

• QA and QC engineers performed several observations of the 1008-VA-L and 1002-GH-L modifica-tion installation activities in the technological part of the plant.

Romeo Bišćan, Igor FifnjaNuklearna elektrarna KrškoVrbina 12, 8270 Krško, [email protected]@nek.si



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• New or revised plant procedures, applicable to modifications 1008-VA-L and 1002-GH-L, were verified and approved by QA.

• QA and QC engineers were involved in review and monitoring of corrective actions implementa-tion through Corrective Action Program.

Quality Assurance involvement in plant modification processes (1008-VA-L and 1002-GH-L) has fulfilled its important role and expectations in achieving overall quality goals and to ensure safe and efficient power plant operation. QA program requirements, as presented in Krško NPP Quality Assurance Plan, were extended to all participants consistent with the importance of their services and scope of supply for nuclear safety.

Keywords: Quality Assurance, QA Plan, Modification, PCFV, PAR


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S6-131

Analysis of Public Opinion Survey “Nuclear Energy – the Present and the Future” (2000–2012)

Technical Museum Zagreb, in co-operation with Hrvatska elektro-privreda, has been organizing high-school students’ visits to the nu-clear power plant Krško since 1989. From schools that have visited the nuclear power plant so far, Zagreb high schools and technical schools are the most represented ones. They filled in the copies of question-naire after watching film about nuclear energy and before visiting nu-clear power plant Krško.The copies of the public opinion survey (questionnaire) have been filled since 2000 and results of analysis of data until 2012 are in this paper. Questions were the same and that makes possible comparison of the re-sults.

Keywords: nuclear power plant, radioactivity, disposal, survey, public acceptance

Irena JakićHrvatska elektroprivreda d.d.Ulica grada Vukovara 37, Zagreb, [email protected]

Renato FilipinTechnical Museum ZagrebSavska cesta 18, Zagreb, [email protected]



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S6-147

Knowledge Loss Risk Assessment in Education and Industry

Knowledge management is based on the idea that the most valuable re-source of some organisation is the knowledge of its people. Organisational performances will depend, among many other things, on how effectively its people can create new knowledge, share knowledge in organisation, and use that knowledge to achieve higher efficiency and the best results. The aim of knowledge management is not necessarily to manage all knowledge, just the knowledge that is most important to the organisa-tion. It is about ensuring that people have the knowledge they need, where and when they need it. Knowledge is derived from information but it is richer and more meaningful than information. In organisational terms, knowledge is generally considered as “knowing how”, or “applied action”. Organisational knowledge is often classified as explicit and tacit knowl-edge. Explicit knowledge can be captured and written down in documents or databases. Tacit knowledge is the knowledge that people carry in their heads and can be difficult to access. Tacit knowledge is considered more valuable because it provides context for people, ideas and experiences.

Knowledge management is discipline consisting of three components: people, processes and technol-ogy. These three components are often compared to the legs of stool – if one is missing, the stool will collapse. However, one component is more important than the others – people. What happens when someone leaves an organisation? Does the organisation feel knowledge loss? According intellectual capi-tal theory organisation will lose not only human capital but also social, structural and relational capital. Determining what happens when these valuable experts leave may help organisation to better understand the impact of knowledge loss and formulate appropriate action in future. Management of knowledge loss is process consisting of three steps: risk assessment, determination of approach for critical knowledge capturing, and monitoring and evaluation. This paper will describe importance of knowledge loss risk assessment in education and industry.

Keywords: knowledge loss, risk assessment, knowledge management

Sanda PlesićUniversity of Zagreb, Faculty of Electrical Engineering and Computing Department of Applied PhysicsUnska 3, 10000 Zagreb, [email protected]

Gonzalo Jimenez VarasTechnical University of Madrid, Nuclear Engineering DepartmentJose Gutierrez Abascal 2, 28006 Madrid, [email protected]


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S6-158

EU Outer Borders and Radiation – An Urgent Need for Standardization, New Detector Technologies and Education Harmonization

It appears that new IAEA transport regulations are “starting to bite” – as now there is a definite requirement for assessments of all relevant radio-nuclides in “non-equilibrium mixes”. It is heard that several shipments that were either delayed significantly or rejected all together in ports in the EU. Things like synthetic rutile, zirconia, copper concentrates, and so on... Devise a NORM Passport for NORM shipments : including a text des-cription of the material and photographs, radionuclide content plus some radiological data (dose rates, gamma spectrum and maybe also something on trigger levels for border monitoring alarms). Produce a guidance booklet for border control authorities. The two sug-gestions are connected, and Ii is suggested that they form a single project of how to equip the EU border contorl facilities. Denial of shipment due to border alarms is an issue that affects industry – that is why it is hoped that they would be interested in being involved. Clearly it is needed to address world-wide transport (i.e. because that’s where most of the NORM comes from). The whole EU outer border customs protocol is to be added to above mentioned monitoring of NORM. The new radiation technologies are used to fulfill the security issues (like stowaways, smuggling, ilicit traf-ficking etc..). All above mentioned forces us to standardize the customs protocols, to upgrade the education of involved workers and to harmonize the issue throughout the EU.Croatia is a front point in the moment because of the very long and „soft“ outer EU border.

Ivica Prlić (1); Marija Surić Mihić (1); Peter Shaw (2); Mladen Hajdinjak (3); Željko Božina (5); Domagoj Kosmina (1),; Zdravko Cerovac (4) 1: Institute for Medical Research and Occupational Health, Ksaverska cesta 2, HR 10000, Zagreb; Republic of Croatia; [email protected], [email protected],[email protected]

2: Public Health England; Duncombe Street Leeds LS1 4PL, United Kingdom, [email protected]

3: Haj-Kom Ltd.; Pavlinski put 5N, HR 10000, Zagreb, Republic of Croatia, [email protected]

4: ALARA Instruments Ltd.; Veslačka 4, HR 10000 Zagreb, Republic of Croatia, [email protected]: Republic of Croatia Ministy of Interior – Police Academy; Avenija G. Šuška 1, HR 10000 Zagreb, Republic of Croatia, [email protected]



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S6-159

NORM – Radionuclide Transfer Studies – A Modern Approach According Directive 2013/59/EURATOM

Regarding NORM, Directive applies in particular to human activities which involve the presence of natural radiation sources that lead to a si-gnificant increase in the exposure of workers or members of the public, in particular the processing of materials with naturally-occurring radio-nuclides (NORM) and the exposure of workers or members of the public to indoor radon, the external exposure from building materials and cases of lasting exposure resulting from the after-effects of an emergency or a past human activity. For areas with long-lasting residual contamination in which the Member State has decided to allow habitation and the resump-tion of social and economic activities, Member States shall ensure, in con-sultation with stakeholders, that arrangements are in place, as necessary, for the ongoing control of exposure with the aim of establishing living conditions that can be considered as normal. For building materials which

are identified by the Member State as being of concern from a radiation protection point of view, Member State shall ensure that, before such materials are placed on the market the activity concentration index I is determined. The index relates to the gamma radiation dose derived throughout the radionuclide transfer studies, in excess of typical outdoor exposure, in a building constructed from a specified building mate-rial. The index applies to the building material, not to its constituents except when those constituents are building materials themselves and are separately assessed as such. The calculation of dose needs to take into account other factors such as density, thickness of the material as well as factors relating to the type of building and the intended use of the material (bulk or superficial).

Ivica PrlićInstitute for medical research and occupational health, , Ksaverska cesta 2, Zagreb, Croatia [email protected]

Mladen HajdinjakHaj-Kom Ltd.; Pavlinski put 5N, HR 10000, Zagreb, Republic of [email protected]


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S6-160

On the Impact to the Human Health from the Fukushima Nuclear Accident

Nuclear energy has long controversial history with hypes of utopian prom-ises and doom fears; from ‘electricity too cheap to meter’ to the certain destruction of the world. Three major nuclear accidents (TMI, Chernobyl and Fukushima) helped only to strengthen mostly negative perception of the public. After TMI and Chernobyl understanding of the nuclear acci-dents impact to human health has considerably improved. This has helped to reduce uncertainty about risk estimates, but too little was achieved re-garding the communication of how absolutely and relatively this risk is in

fact small. Now with Fukushima there is new potential for even better understanding of the realistic scope of the total nuclear accident impact. Based on the experience with previous accidents obviously there are significant obstacles to achieve that.This paper is presenting current understanding of the impact on the human health from the Fukushima nuclear accident. Focus is given also to the uncertainties, perception, and relative prospective to other non radiation related risks. It is clear from data and assessments that immediate radiation related risk as well as expected long term health impact is even with conservative approach negligible and that it will be most probable impossible to determine it with health monitoring and epidemiological study. However, because of existing perception, lack of thrust and better approach many practical reactions are going to perpetuate negative picture, about radiation and nuclear energy, and perhaps cause some additional psychosocial risk. It is troubling that this induced risk is substantial and it could have been mainly pre-vented by timely applying experience form previous major accidents thru education, communication and transparency. Because of historical and local circumstance, communication difficulties and lack of sufficient understanding of uncertainties and magnitudes it seems that Fukushima nuclear accident will not be much better case than Chernobyl. This is unfortunate because of potential to improve on existing experience and clear confusion between uncertainty and importance.Considering lack of proper practical learning from previous nuclear accidents it seems that there is no easy way to treat Fukushima accident according to the objective scale of impact. Instead, majority of measures have to account for strong negative public perception and act as like impact is significant. This than in fact is inducing and perpetuating psychosocial problems and eventually increasing real negative heath impacts. Without serious effort to communicate early enough low importance of nuclear accident impact and real meaning of existing scientific uncertainty it will be impossible to change this even after Fukushima accident. This is unfortunate for post Fukushima recovering as well for any future response in the area of nuclear safety, emergency management and new nuclear development.

Keywords: Fukushima nuclear accident, human health impact, uncertainty, perception, risk communication

Zdenko ŠimićUniversity of Zagreb Faculty of Electrical Engineering and Computing Unska 3, 10000 Zagreb, [email protected]


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Session 7 Regulatory Practice and Emergency Preparedness

S7-110 N. NovoselEU Directives in the Field of Radiological and Nuclear Safety and Their Transposition in Croatian Legislation 90

S7-119 G. PogačićCyber Security in Nuclear Power Plants – U.S. NRC Regulatory Guide 5.71 91

S7-127 Z. BazsÓImpact of TEPCO Fukushima Dai-ichi accident on severe accident management in the Slovak Republic 92

S7-128 V. Kuznetsov, Z. Drace, V. LysakovMajor Findings of IAEA/INPRO Activity on Legal and Institutional Issues for Transportable Nuclear Power Plants 93

S7-143 S. Medaković, R. BanovAssessment of NPP Krško accident impact on the population by means of the nonlinear statistical model using the RODOS software package 94

S7-155 J. Vuković, D. Konjarek, D. GrgićRadiation doses estimation for hypothetical NPP Krško accidents without and with PCFV using RASCAL software 95

S7-164 I. Bašić, M. Kim, P. Hughes, B. K. Lim, F. D’auria, M. L. LouisIAEA Review for Gap Analysis of Safety Analysis Capability 96

S7-170 D. Konjarek, T. Bajs, D. ŠinkaPreliminary radiation doses assessment for NPP Krško SBO sequence 98

Session 7 - Regulatory Practice and Emergency Preparedness (RPEP)90

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Nevenka NovoselState Office for Radiological and Nuclear SafetyFrankopanska 11, 10 000 Zagreb, [email protected]

S7-110

EU Directives in the Field of Radiological and Nuclear Safety and Their Transposition in Croatian Legislation

In the process of accession to European Union, Croatia was obliged to transpose EU directives in the field of radiological and nuclear safety in Croatian legislation. These directives are:

• Council Directive 89/618/Euratom of 27 November 1989 on in-forming the general public about health protection measures to be applied and steps to be taken in the event of a radiological emer-gency,

• Council Directive 90/641/Euratom of 4 December 1990 on the operational protection of outside workers exposed to the risk of ionizing radiation during their activities in controlled areas,

• Council Directive 96/29/Euratom of 13 May 1996 laying down basic safety standards for the pro-tection of the health of workers and the general public against the dangers arising from ionizing radiation,

• Council Directive 97/43/Euratom of 30 June 1997 on health protection of individuals against the dangers of ionizing radiation in relation to medical exposure, and repealing Directive 84/466/Eur-atom,

• Council Directive 2003/122/Euratom of 22 December 2003 on the control of high-activity sealed radioactive sources and orphan sources,

• Council Directive 2006/117/Euratom of 20 November 2006 on the supervision and control of ship-ments of radioactive waste and spent fuel,

• Council Directive 2009/71/Euratom of 25 June 2009 establishing a Community framework for the nuclear safety of nuclear installations and

• Council Directive 2011/70/Euratom of 19 July 2011 establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste.

In this paper aforementioned directives will be presented, with the accent on Croatian legislation in which transposition was done, as well as new revisions of those directives and foreseen revisions in the near future.

Keywords: European Union, directives, transposition, radiological safety, nuclear safety, legislation

Session 7 - Regulatory Practice and Emergency Preparedness (RPEP) 91


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S7-119

Cyber Security in Nuclear Power Plants – U.S. NRC Regulatory Guide 5.71

We have already made a big step into new millennia and with it there is no more dilemma about presence of computers and internet in our lives. Almost all modern facilities struggle with this new dimension of informa-tion flow and how to use it to their best interest. But there is also the other side of the coin – the security threat. For nuclear power plants this threat poses even greater risk. In addition to protecting their trade secrets, per-sonal data or other common targets of cyber attacks, nuclear power plants

need to protect their digital computers, communication systems and networks up to and including the design basis threat (DBT). As stated in U.S. Nuclear Regulatory Commission (NRC) Regulatory Com-mission Regulations, Title 10, Code of Federal Regulations (CFR), section 73.1, “Purpose and Scope” this includes protection against acts of radiological sabotage and prevention of the theft or diversion of special nuclear material.The main purpose of this paper is to explore the NRC Regulatory Guide (RG) 5.71 and its guidance in implementing cyber security requirements stated in NRC 10 CFR, section 73.54, “Protection of Digital Computer and Communication Systems and Networks”. In particular, this section requires protection of digital computers, communication systems and networks associated with the following categories of functions:

• safety-related and important-to-safety functions,• security functions,• emergency preparedness functions, including offsite communication, and• support systems and equipment which, if compromised, would adversely impact safety, security, or

emergency preparedness functions.

This section requires protection of such systems and networks from those cyber attacks that would act to modify, destroy, or compromise the integrity or confidentiality of data or software; deny access to sys-tems, services or data; and impact the operation of systems, networks, and equipment.This paper will also present some of experiences from implementing these regulations in NPP Krško.

Keywords: cyber security, regulatory guide, NPP Krško, 10 CFR 73.54

Goran PogačićNPP Krško, Engineering Services, Process InformaticsVrbina 12, 8270 Krško, [email protected]


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S7-127

Impact of TEPCO Fukushima Dai-ichi accident on severe accident management in the Slovak Republic

The findings of Periodic Safety Review (PSR) performed recently accord-ing to the internationally acceptable methodology for Nuclear Power Plant (NPP) units operated in the Slovak Republic resulted in a correc-tive plan of measures with the aim to eliminate shortcomings identified in the frame of PSR and to adopt safety improvements as well. One group of measures concerning the area of Severe Accident Management (SAM) has been incorporated into the renewed operating license issued by the Nuclear Regulatory Authority of the Slovak Republic (UJD SR).

Strengthening of NPP resistance against events with low probability of occurrence but potentially with very severe consequences has been continued in the Slovak Republic following the accident in the Fuku-shima Daiichi Nuclear Power Plant in 2011. The operator’s Action Plan (AcP) based on lessons learned from the events at the NPP Fukushima Daiichi is the first part of operator activities concerning urgent corrective measures to prevent reoccurrence of similar events.The National Action Plan of Slovak Republic (NAcP SR) based on the AcP, submitted by the licence hold-er of units operated in the Slovak Republic to the UJD SR, reflects on all related European Nuclear Safety Regulators Group (ENSREG) recommendations. The implementation of all intended plant modifications, as the second part of operator activities in this field, has been supervised by the UJD SR, starting with the review of documents on the given subject and ending with inspections of the site.The paper provides an overview of history and the progress of the activities performed recently by the operator to meet the conditions stated in the renewed operating license in the area of Severe Accident Management (SAM) and AcP mentioned above, with the objective to upgrade the nuclear safety of nu-clear installations in the Slovak Republic in case of severe accidents.

Keywords: nuclear power plant, severe accident management, action plan, regulatory body, nuclear safety

Zoltán BazsÓ Urad jadroveho dozoru Slovenskej republikyOkružná 5, 918 64 Trnava, Slovak [email protected]



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S7-128

Major Findings of IAEA/INPRO Activity on Legal and Institutional Issues for Transportable Nuclear Power Plants

The IAEA’s International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000. INPRO cooperates with Mem-ber States to ensure that sustainable nuclear energy is available to help meet the energy needs of the 21st century. INPRO is part of the integrated services of the IAEA provided to Member States considering initial devel-opment or expansion of nuclear energy programmes. INPRO study on legal and institutional issues of Transportable Nuclear Power Plants (TNPPs) has been finalized in 2013. A transportable nuclear power plant (TNPP) is a factory-manufactured, movable nuclear power plant, which when fuelled is capable of producing final energy products such as electricity and heat. Transportable nuclear power plants are not designed to operate during transportation. This publication highlights the potential benefits of TNPPs, describes the legal and institutional issues for their deployment in countries other than the country of origin, reveals challenges that might be faced in their deployment, and outlines pathways for resolution of the identified issues and challenges in the short and long terms. It is addressed to senior legal, regulatory and technical officers in Member States planning to embark on a nuclear power programme or to

expand an existing one by considering the introduction of a TNPP.

Keywords: transportable nuclear power plant, factory fuelled reactor, international legal framework, nu-clear infrastructure, safety, security, safeguards

Vladimir Kuznetsov, Zoran Drace,International Atomic Energy AgencyVienna International Centre, PO Box 100, 1400 Vienna, [email protected] , [email protected]

Viacheslav LysakovInternational Atomic Energy AgencyVienna International Centre, PO Box 100, 1400 Vienna, [email protected]


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S7-143

Assessment of NPP Krško accident impact on the population by means of the nonlinear statistical model using the RODOS software package

It was our intention, in this study, to assess the impact on the population in terms of received doses for different routes of exposure after a severe nuclear accident, and to determine the optimal planning distance param-eter in the area of application of urgent measures of protection and rescue. The primary objective was to determine the model of exposure in terms of the received effective dose and thyroid dose as well as the received bone marrow dose which allows statistical observation of the calculation results and the determination of appropriate distribution. The primary analysis was performed by repeatedly launching the computing system RODOS with modified input data in consideration of wind speed and the atmos-pheric stability class. Model exposure is based on the statistical analysis of the data collected. This paper shows the results of repeated launching

of the RODOS system and the statistical processing of the results. Described in detail it approaches the analysis, defining input data in form of PSA source terms, further to this, the calculation model based on available models of the dispersion of radioactive clouds in the RODOS system and the method and results of statistical processing of data collected in a form that allows easy evaluation of consequences in relation to the distance from the NPP Krško. The method of analysis was carried out at the premises of the State Office for Radiological and Nuclear Safety (SORNS) using the available computational resources of the SORNS.

Keywords: Exposure model, effective dose, thyroid, bone marrow, random field, RODOS

Saša MedakovićState Office for Radiological and Nuclear SafetyFrankopanska 2, Zagreb, [email protected]

Reni BanovFaculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected]



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Josip Vuković, Damir KonjarekEnconet d.o.o.Miramarska 20, Zagreb, [email protected], [email protected]

Davor GrgićUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected]

S7-155

Radiation doses estimation for hypothetical NPP Krško accidents without and with PCFV using RASCAL software

Calculation is done using Source Term to Dose module of RASCAL (Ra-diological Assessment System Consequence Analysis) software to esti-mate projected radiation doses from a radioactive plume to the environ-ment. Utilizing this module, it is possible to do preliminary assessment of consequences to the environment in case of adverse reactor conditions or releases from other objects containing radioactive materials before the emergency situation has happened or in the early phase. RASCAL is sim-ple, easy to use, fast-running tool able to provide initial estimate of radio-logical consequences of nuclear accidents.Upon entering rather limited amount of input parameters for the Krško NPP, mostly key plant parameters, time dependent source term calcula-tion is executed to determine radioactive inventory release rates for differ-ent plant conditions, release paths and availability of protective measures. These rates given for each radionuclide as a function of time are used as an input to atmospheric dispersion and transport model. Together with re-

lease rates, meteorological conditions dataset serve as input to determine the behavior of the radioactive releases that is plume in the atmosphere. So as an output, RASCAL produces a “dispersion envelope” of radionuclide concentrations in the atmosphere. These concentrations of radionuclides in the atmosphere are further used for estimating the doses to the environment and the public downwind the release point. Throughout this paper, dose assessment is performed for two distances, close-in distance and distance out to 40 km from the source, for hypothetical NPP Krško accidents without and with Passive Contain-ment Filtered Vent (PCFV) system used. Obvious difference is related to released radioactivity of Iodine isotopes. Results of radioactive effluents deposition in the environment are displayed via various doze parameters, radionuclide concentrations and materials exposure rates in this particular case.

Keywords: radiation doses, RASCAL, Krško NPP, release rates, iodine, PCFV


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S7-164

IAEA Review for Gap Analysis of Safety Analysis Capability

The IAEA Asian Nuclear Safety Network (ANSN) was launched in 2002 in the framework of the Extra Budgetary Programme (EBP) on the Safety of Nuclear Installations in the South East Asia, Pacific and Far East Coun-tries. The main objective is to strengthen and expand human and ad-vanced Information Technology (IT) network to pool, analyse and share nuclear safety knowledge and practical experience for peaceful uses in this region. Under the ANSN framework, a technical group on Safety Analysis (SATG) was established in 2004 aimed to providing a forum for the ex-change of experience in the following areas of safety analysis:

• To provide a forum for an exchange of experience in the area of safety analysis,

• To maintain and improve the knowledge on safety analysis method,• To enhance the utilization of computer codes,• To pool and analyse the issues related with safety analysis of re-

search reactor, and• To facilitate mutual interested on safety analysis among member

countries.

A sustainable and successful nuclear energy programme requires a strong technical infrastructure, including a workforce made up of highly special-

ized and well-educated professionals. A significant portion of this technical capacity must be dedicated to safety — especially to safety analysis — as only then can it serve as the basis for making the right deci-sions during the planning, licensing, construction and operation of new nuclear facilities. In this regard, the IAEA has provided ANSN member countries with comprehensive training opportunities for capacity building in safety analysis. Nevertheless, the SATG recognizes that it is difficult to achieve harmoniza-tion in this area among all member countries because of their different competency levels. Therefore, it is necessary to quickly identify the most obvious gaps in safety analysis capability and then to use existing resources to begin to fill those gaps.The goal of this Expert Mission (EM) for gap finding service is to facilitate improvement of nuclear safety in the participating host organization and host member countries. To achieve this goal, the EM is to es-tablish a process of discussion and comparison of gap findings, which will lead to sharing of information,

Ivica BašićAPoSS d.o.o.Zabok, [email protected]

Manwoong Kim, Peter Hughes, B-K LimInternational Atomic Energy Agency (IAEA)Vienna, Austria

Francesco D’auriaUNIPIc/o DIMNO – L.go L. Lazzarino 156100 Pisa, Italy

Vidard Michel LouisNumericableSaint Genis Laval69230 Saint Genis Laval, France



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experience, strengths and weaknesses among the participants, and foster regional cooperation to improve the weaknesses and improve safety generally.The pilot mission was conducted from 28 October to 1 November for one week at the National Nuclear Agency (BATAN) in Indonesia by the mission team formulated with 6 international experts who have considerable knowledge and experience in the field of safety analysis such as the deterministic safety analysis (DSA) and probabilistic safety analysis (PSA). Some comments and recommendations were given to BATAN management to support the establishment and maintenance of safety analysis capability and human resource, organizational and training aspects. Those aspects are important as a measure of the progress being made and an indicator of areas in SATG within the framework of the Extra-budgetary Programme on the Safety of Nuclear Installations in Southeast Asia, the Pacific, and Far East Countries (the EBP-Asia) or other cooperation programme, such as the IAEA Technical Cooperation programme.Provided in 2013 the Review of Gap Analysis for BATAN (Indonesian Nuclear Safety Regulatory Body) could be good reference for all other newcomer countries which started or plans nuclear power plant installation.

Keywords: safety analyses capability, nuclear power plants, review of gap analysis


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S7-170

Preliminary Radiation Doses Assessment for NPP Krško SBO Sequence

Accidents at nuclear power plants may require urgent measures to pre-vent deterministic effects and reduce stochastic consequences. The need for fast decision making is restricted with great uncertainty of what is known about the situation in the early phase of an accident. In particular, magnitude, beginning and duration of a release are subject to consider-able uncertainty and meteorological information might be unavailable or vague. This calls for simplistic and robust decision criteria which is de-fined in emergency preparedness zones.Recent developments on international level, especially after Fukushima accident emphasized need to have harmonized emergency preparedness zones in case of close proximity of nuclear power plants to the national

borders. Example of such a case is Nuclear Power Plant Krško (NEK) that is located just 10 km from Cro-atian-Slovenian border. Emergency zones in Croatia and Slovenia do not have adequate level of harmoni-zation. Croatian State Office for Radiological and Nuclear Safety (DZRNS – Državni zavod za radiološku i nuklearnu sigurnost) initiated an activity to gather background information based on which harmoniza-tion of emergency preparedness planning zones could be done between Croatia and Slovenia.In order to achieve this, use of NEK PSA (Probabilistic Safety Assessment) Level II, LERF (Large, early release frequency) and Release Frequency (RF) results has been explored. Additionally, impact of recent modification PCFVS (Passive Containment Filtered Vent System) has been investigated. Release category corresponding to the SBO (Station BlackOut) was selected to investigate the influence of PCFVS modi-fication that enables controlled and filtered release of containment structure. IAEA available code IN-TERRAS (INTERnational Radiological Assessment System) module ST-Dose (Source Term to Dose) was used to estimate projected radiation doses from a radioactive plume to the environment. With this mod-ule it is possible to make preliminary assessment of consequences to the environment in case of adverse reactor conditions or releases from radioactive materials. Its intended usage is for quick assessment of consequences of radiological release before and during initial phases of nuclear accident when radiologi-cal measurements is still not possible or is limited. This paper represent preliminary finding as the conservative boundary conditions in calculation yield doses close to intervention levels.

Keywords: radiation doses, INTERRAS, Krško NPP, SBO, release rates, iodine, LERF, filtered venting

Damir Konjarek, Tomislav Bajs, Davor ŠinkaEnconet d.o.o.Miramarska 20, HR-10000 Zagreb, [email protected]@[email protected]


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Session 8 Reactor Physics and Nuclear Fuel Cycle

S8-100 M. Matijević, D. Pevec, K. TrontlSCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations 100

S8-120 J. Haščik, G. Farkas, B. Vrban, J. Lüley, P. UrbanPower Density Determination in the VVER 440 Reactor by the Transport Codes MCNP 5 and SCALE 6 101

S8-130 M. Kromar, B. KurinčićInfluence of the Finer Radial Burnup Nodalization on the Pin Power Distribution in the PWR core 102

S8-144 M. Božič, M. Kromar, B. KurinčićFuel Reloading Strategies in a Hypothetical NPP Krško Forced Outage 103

S8-152 D. Grgić, S. Šadek, V. Benčik, D. KonjarekDecay Heat Calculation for Spent Fuel Pool Application 104

S8-154 R. Ječmenica, M. Matijević, D. GrgićFuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs 105

S8-156 D. Zhang, F. RahnemaEfficiency and Accuracy of the Incident Flux Response Expansion Method for LaBr3 Detector Pulse Height Spectrum Calculation 106

Session 8 - Reactor Physics and Nuclear Fuel Cycle (RPNFC)100

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S8-100

SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

The capabilities and limitations of SCALE6/MAVRIC hybrid determinis-tic-stochastic shielding methodology (CADIS and FW-CADIS) are dem-onstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by prepa-ration of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR pa-rameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in

this paper was determination of neutron-gamma dose rate distribution (radiation field) over large por-tions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radia-tion included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical pa-rameters of deterministic module plays important role for computer memory management. We investi-gated the possibility of using deterministic module (memory intense) with broad group library v7_27n19g opposed to fine group library v7_200n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with CADIS and also analog MC simulations, the FW-CADIS drastically improved MC dose rate calculations in quality as well in quantity. Large shielding problems such as portions and complete PWR facility require not only extensive computational resources but also understanding of the underlying physics, which is inevitable in interpreting results of hybrid deterministic-stochastic methodology.

Keywords: Monte Carlo, SCALE6, shielding, hybrid methodology, FW-CADIS, CADIS, variance reduction

Mario Matijević, Dubravko Pevec, Krešimir TrontlUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, 10000 Zagreb, [email protected], [email protected], [email protected]

Session 8 - Reactor Physics and Nuclear Fuel Cycle (RPNFC) 101


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S8-120

Power Density Determination in the VVER 440 Reactor by the Transport Codes MCNP 5 and SCALE 6.

This paper deals with the verification of neutron power density in the active core (AC) of Mochovce units in the selected campaign using codes MCNP5 and SCALE 6.1.1. This analysis serves as an independent assessment of power density spatial distribution in the reactor core after loading the new type fuel of assemblies with an av-erage enrichment of 4.87% 235U. The codes NJOY and SCALE 6.1.1 responsible for the temperature dependent cross-sections librar-ies generation and determination of fuel radionuclide composition as a function of operational history were used. The analysis of the impact of reactor power level uncertainty (decrease) to the calculated isotopic composition and the multiplication abilities of loaded as-semblies was also carried out. The paper gives a brief description of the geometrical and material models used in the calculations. The calculated spatial distribution of power density correlates with the power density distribution determined by on site power monitoring system. The calculation demonstrates the significant impact of the fuel assemblies with higher enrichment to the power density distri-bution in their vicinity.

Keywords: power density, VVER 440 reactor, radionuclide composition, uncertainty, fuel assembly, cam-paign, enrichment

Jan Haščik, Gabriel Farkas, Branislav Vrban, jakub LüleySlovak University of Technology, Institute of Nuclear and Physical Engineering Iľkovičova 3, 81219 Bratislava, Slovak [email protected], [email protected], [email protected], [email protected]

Perter UrbanSlovenské elektrárne a.s NPP MochovceJE Mochovce, 93539 Mochovce, Slovak [email protected]


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S8-130

Influence of the Finer Radial Burnup Nodalization on the Pin Power Distribution in the PWR core

Influence of the finer burnup nodalization on the core reactivity and pow-er distribution is analyzed in a typical NPP Krško 18-month cycle with the CORD-2 system. A model with 2 x 2 nodes per assembly is compared with a calculational sequence using just one node per assembly. Significant dif-ferences are observed. Finer burnup nodalization more reliably describes neutron leakage from the reactor core. Higher calculated leakage resulted in a 28 ppm lower full power critical boron concentration at the end of the cycle. Up to 2.5 % differences in control rods worth were observed. Finer burnup nodalization predicts lower pin by pin burnup gradients for the assemblies at the core periphery, where the burnup gradients are the highest. At the core inside locations this mechanism can be distracted by

the global power shift caused by the higher neutron leakage in peripheral assemblies.

Keywords: PWR, burnup, pin by pin power, CORD-2

Marjan Kromar‘Jožef Stefan’ InstituteJamova 39, 1001 Ljubljana, [email protected]

Bojan KurinčičNuclear Power Plant KrškoVrbina 12, 8270 Krško, [email protected]



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S8-144

Fuel Reloading Strategies in a Hypothetical NPP Krško Forced Outage

Reload strategies in a premature unscheduled outage (forced outage) of the NPP Krško are investigated. It is assumed, that baffle jetting could cause degradation of fuel assemblies at specific near baffle locations, which would consequently require prompt fuel replacement. Loading pat-terns after 6 and 12 month of operation in a typical 18 month fuel cycle are developed satisfying all nuclear design criteria. They are determined with the ROSA optimization code system, while the important reactor core parameters such as power peaking factors, temperature coefficients, rods worth are calculated and verified with the CORD-2 system.

Keywords: PWR, forced shutdown, loading pattern

Matjaž Božič, Bojan KurinčičNuclear Power Plant KrškoVrbina 12, 8270 Krško, [email protected]

Marjan Kromar‘Jožef Stefan’ InstituteJamova 39, 1001 Ljubljana, [email protected]


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S8-152

Decay Heat Calculation for Spent Fuel Pool Application

The automatic procedure was developed for fuel assembly decay heat cal-culation based on PARCS 3D burnup calculation for fuel cycle depletion, and ORIGEN 2.1 calculation during both depletion and fuel cooling. Us-ing appropriate pre-processor and post-processor codes it is possible to calculate fuel assembly decay heat loads for all fuel assemblies discharged from reactor. Simple graphical application is then used to distribute fuel assemblies within fuel pool and to calculate any fuel assembly, SFP rack, or whole pool heat load at arbitrary time. The application can be used for overview of fuel assembly burnups, cooling times or decay heats. Based on given date it is possible to calculate whole pool heat load and time to boiling or time to assembly uncovery using simple mass and energy balances. Calculated heat loads can be input to more detailed thermal-hydraulics calculations of spent fuel pool. The demonstration calculation was performed for NPP Krsko spent fuel pool.

Keywords: spent fuel pool, decay heat calculation, ORIGEN, graphical representation of data

Davor Grgić, Siniša šadek, Vesna BenčikUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected], [email protected], [email protected],

Damir KonjarekEnconet d.o.o.Miramarska 20, Zagreb, [email protected]



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S8-154

Fuel Depletion Modeling of Reconstituted NEK Fuel Assembly Using Lattice Cell Programs

This paper presents a transport fuel depletion calculations of one of the four NPP Krško reconstituted fuel elements obtained using lattice cell programs (FA2D and TRITON/NEWT). These elements were foreseen for use in three fuel cycles, but after the second cycle seven boundary fuel rods were replaced with steel rods, of equal diameter, to preclude potential fuel rod damage due to baffle jetting. This replacement intro-duces additional asymmetry in the 16x16 fuel matrix geometry because all seven steel rods are located along two external boundaries. The reconsti-tuted fuel elements after the second cycle are placed at the core periphery (close to the reflector-baffle region where repeated baffle jetting damage is expected). The depletion calculation is done up to 60 GWd/MtU using nominal power of 40.5 W/gU. The “restart” option with material change

was used after the fuel assembly reconstitution at burnup of 26 GWd/MtU (actual burnup acquired by fuel assembly in NPP Krsko cycle 26). We observed significant change of pin-power distribution so ad-ditional Monte Carlo calculations of the modified element were done at 26 GWd/MtU using KENO-VI and MCNP5 programs. Isotopic composition of the fuel pins at the given burnup, needed in Monte Carlo calculation, is obtained from TRITON/NEWT run. The comparison in calculated fuel assembly homog-enized properties as well as in 2D spatial distribution (pin powers) was provided at the end of paper.

Keywords: fuel assembly reconstitution, 2D transport calculation, depletion, pin power distribution

Radomir Ječmenica, Mario Matijević, Davor GrgićUniversity of Zagreb, Faculty of Electrical Engineering and ComputingUnska 3, Zagreb, [email protected], [email protected], [email protected]


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S8-156

Efficiency and Accuracy of the Incident Flux Response Expansion Method for LaBr3 Detector Pulse Height Spectrum Calculation

An incident flux response expansion (IFLEX) method has recently been developed to compute the pulse height response of radiation detectors due to an incident photon radiation on-the-fly. The IFLEX method for modelling radiation detectors is based on first expanding the incident photon energy distribution on the detector window in terms of a set of known expansion functions (B-spline) and then superimposing pre-computed a set of response functions to construct the pulse height spec-tra. The method has been previously benchmarked against MCNP for a CsI scintillator with mono-directional incident sources. In this work, the

IFLEX method is extended to handle incident flux with discrete phase space distribution obtained from deterministic methods such as the discrete ordinates method. In this case, the phase space distribution at the detector window is expanded by a tensor product of B-spline, Legendre polynomials and discrete real spherical harmonics. A test problem composed of a LaBr3 scintillator and an incident flux representing incident fluxes from typical cargo containers is used to evaluate the accuracy and computational efficien-cy of the method. It is found that the pulse height response predicted by the IFLEX method agrees very well with that directly computed by direct continuous energy Monte Carlo method. The average relative differences between the methods are less than 2%. The discrepancies are within three standard deviations of the Monte Carlo reference solutions. The IFLEX method is found to be significantly (about 7 orders of magnitude) faster than the Monte Carlo method.

Keywords: detector calculation, stochastic, incident flux, pulse height

Dingkang Zhang, Farzad RahnemaGeorgia Institute of Technology770 State Street, Atlanta, GA 30332, USA [email protected], [email protected]



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Index of Authors

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